SRS Liquid Waste Facilities Performance Assessment SRR ...• Special Analyses (SAs) • PA revisions This program Implementation Plan is prepared and updated annually and submitted

SRS Liquid Waste Facilities Performance Assessment SRR-CWDA-2017-00096 Maintenance Program – FY2018 Implementation Plan Revision 0 January 2018

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TABLE OF CONTENTS

TABLE OF CONTENTS ............................................................................................................. 2

LIST OF FIGURES ...................................................................................................................... 4

ACRONYMS/ABREVIATIONS ................................................................................................. 5

1.0 INTRODUCTION ................................................................................................................. 7

2.0 SALTSTONE DISPOSAL FACILITY ............................................................................. 12

2.1 Saltstone Disposal Facility Performance Assessment Annual Maintenance Activities ... 12 2.1.1 Maintain Saltstone Disposal Facility Performance Assessment Control through

Unreviewed Waste Management Question Process ..............................................12 2.1.2 Prepare Annual Performance Assessment Maintenance Program Implementation

Plan ........................................................................................................................12 2.1.3 Provide General Technical Support on Saltstone Disposal Facility Performance

Assessment Issues...................................................................................................13 2.1.4 Develop and Maintain Performance Assessment Model Archive and Revision

Control ...................................................................................................................13 2.1.5 Conduct Annual Saltstone Disposal Facility Performance Assessment Validation ...13 2.1.6 Maintain Saltstone Disposal Facility Performance Assessment Closure Plan ..........14 2.1.7 Maintain Saltstone Disposal Facility Performance Assessment Monitoring Plan ....14

2.2 Saltstone Disposal Facility Performance Assessment Development/Revisions ............... 15 2.2.1 Prepare Out-Year Saltstone Disposal Facility Performance Assessment

Revisions ................................................................................................................15 2.2.2 Saltstone Disposal Facility Special Analyses .............................................................15

2.3 Saltstone Disposal Facility Performance Assessment Testing & Research Activities ..... 16 2.3.1 Critical Property Testing ............................................................................................18 2.3.2 Ongoing Studies .........................................................................................................30 2.3.3 Emplaced Saltstone Sampling and Characterization .................................................38 2.3.4 To Be Determined Out-Year Testing ..........................................................................50

3.0 F-AREA AND H-AREA TANK FARMS .......................................................................... 51

3.1 Tank Farm Performance Assessment Annual Maintenance Activities ............................. 51 3.1.1 Maintain Tank Farm Performance Assessment Control through Unreviewed

Waste Management Question Process ...................................................................51 3.1.2 Prepare Annual Performance Assessment Maintenance Program Implementation

Plan ........................................................................................................................51 3.1.3 Provide General Technical Support on Tank Farm Performance Assessment

Issues ......................................................................................................................52 3.1.4 Develop and Maintain Performance Assessment Model Archive and Revision

Control ...................................................................................................................52 3.2 Tank Farm Performance Assessment Development/Revisions ........................................ 52

3.2.1 Prepare Out-Year F-Area Tank Farm Performance Assessment Revisions ..............53 3.2.2 Prepare Out-Year H-Area Tank Farm Performance Assessment Revisions ..............54 3.2.3 Tank Farm Special Analyses ......................................................................................55


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3.3 Tank Farm Performance Assessment Testing & Research Activities .............................. 55 3.3.1 Tank Residual Characterization .................................................................................56 3.3.2 To Be Determined Out-Year Testing ..........................................................................59

4.0 REFERENCES .................................................................................................................... 61

APPENDIX A .............................................................................................................................. 72

APPENDIX B .............................................................................................................................. 77

APPENDIX C .............................................................................................................................. 99


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LIST OF FIGURES

Figure 1.0-1: Disposal Document Activities for Saltstone Disposal Facility ................................ 9 Figure 1.0-2: Closure Document Development Activities for the Tank Farms ........................... 11 Figure 2.3-1: Saltstone Research, Development, and Testing Program Elements ...................... 17 Figure 2.3-2: Critical Property Testing Strategy.......................................................................... 19 Figure 2.3-3: Leachate Concentrations for Tc-spiked DLM Sample .......................................... 28 Figure 2.3-4: SHC and pH Plotted as a Function of Pore Volumes Exchanged for Tc-spiked

DLM Sample ...................................................................................................... 28 Figure 2.3-5: Cumulative % Tc-99 and Nitrate Leached from Tc-spiked DLM Sample ............ 29 Figure 2.3-6: Leachate pH and Tc Concentrations for Tc-spiked DLM Sample......................... 29 Figure 2.3-7: Ongoing Studies ..................................................................................................... 32 Figure 2.3-8: Emplaced Saltstone Testing Strategy (Detailed in SRR-SPT-2012-00049) .......... 41 Figure 2.3-9: SHC Versus Pore Volumes Exchanged for SDU Cell 2A Cores ........................... 47 Figure 2.3-10: Leachate Concentration Data for SDU Cell 2A (SDU2A-0931-C-1-U-5) .......... 47 Figure 2.3-11: Leachate Concentration Data for SDU Cell 2A (SDU2A-0931-C-1-U-2) .......... 48 Figure 2.3-12: Leachate pH and Tc Concentrations for SDU Cell 2A DLM Samples ............... 48 Figure 3.3-1: Photograph of Testing Equipment Prior to Installation into the SRNL Shielded

Cell ..................................................................................................................... 58

LIST OF TABLES

Table 2.3-1 DLM-Derived Data for SDU Cell 2A Samples. ....................................................... 46


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ACRONYMS/ABREVIATIONS

ASR Alkali-Silica Reactivity ADMP Advance Design Mixer Pump AGW Artificial Groundwater ALARA As Low As Reasonably Achievable ASTM American Society for Testing and Materials BFS Blast Furnace Slag CA Composite Analysis CAB Citizens Advisory Board CBP Cementitious Barriers Partnership CFR Code of Federal Regulations CY Calendar Year DAS Disposal Authorization Statement DLM Dynamic Leaching Method DO Dissolved Oxygen DOE U.S. Department of Energy DOE-SR U.S. Department of Energy Savannah River Operations Office Eh Measure of Reduction (or Oxidation) Potential EPA U.S. Environmental Protection Agency FEPs Features, Events, and Processes FFA Federal Facility Agreement FTF F-Area Tank Farm FY Fiscal Year GCL Geosynthetic Clay Liner HDPE High Density Polyethylene HRR Highly Radioactive Radionuclide HTF H-Area Tank Farm ICP-MS Inductively Coupled Plasma – Mass Spectrometry Kd Distribution Coefficient LFRG Low-Level Waste Disposal Facility Federal Review Group LW Liquid Waste MA Monitoring Area MCC Moisture Characteristic Curve MEP Maximum Extent Practical MF Monitoring Factor MOP Member of the Public N/A Not Applicable NDAA National Defense Authorization Act NRC U.S. Nuclear Regulatory Commission


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PA Performance Assessment PARC Performance Assessment Review Committee pH Measure of Acidity or Alkalinity of a Solution PNNL Pacific Northwest National Laboratory POC Point of Compliance RAI Request for Additional Information Rd Distribution Ratio RH Relative Humidity SA Special Analysis SAP Sampling and Analyses Plan SCDHEC South Carolina Department of Health and Environmental Control SDF Saltstone Disposal Facility SDU Saltstone Disposal Unit SPF Saltstone Production Facility SHC Saturated Hydraulic Conductivity SREL Savannah River Ecology Laboratory SRNL Savannah River National Laboratory SRNS Savannah River Nuclear Solutions, LLC SRR Savannah River Remediation LLC SRS Savannah River Site TER Technical Evaluation Report TTQAP Task Technical and Quality Assurance Plan TRR Technical Review Report UWMQ Unreviewed Waste Management Question UWMQE Unreviewed Waste Management Question Evaluation WAC Waste Acceptance Criteria WDA Waste Disposal Authority WRM Waste Release Model


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1.0 INTRODUCTION

The U.S. Department of Energy (DOE), through Manual 435.1-1, Radioactive Waste Management Manual, and associated guidance, requires the ongoing maintenance of all Performance Assessments (PAs) and Composite Analyses (CAs). Because PA and CA results, in part, are based on data that is uncertain due to utilization of projected conditions thousands of years into the future, a maintenance program is needed to continue to reduce uncertainty in the inputs and assumptions, providing greater confidence in the results of the analyses and in the long-term plans for public and environmental protection. Additionally, a disciplined process to address potential changes in disposal and/or closure operations (e.g., change in disposal unit design, new residual material characterization) is needed to ensure that proposed changes do not adversely affect conclusions reached using PA results.

The PAs for the Savannah River Site (SRS) Saltstone Disposal Facility (SDF), F-Area Tank Farm (FTF), and H-Area Tank Farm (HTF) are managed by Savannah River Remediation LLC (SRR) for the DOE. [SRR-CWDA-2009-00017, SRS-REG-2007-00002, SRR-CWDA-2010-00128] These PAs assess the calculated dose impact on a future, hypothetical member of the public (MOP) and an inadvertent human intruder, as well as environmental impacts from the respective facilities after final closure as required by DOE Order 435.1, Change 1. In addition, the PAs are used to support demonstration of compliance with pertinent requirements of the Ronald W. Reagan National Defense Authorization Act (NDAA) for Fiscal Year 2005, Section 3116 (hereinafter referred to as NDAA Section 3116). The Savannah River Site DOE 435.1 Composite Analysis (hereinafter referred to as: SRS CA) is a management tool required to assist DOE in assessing the possible impacts on the public and environment from multiple sources of legacy radioactive material at a DOE site (e.g., SRS) in order to determine where DOE may need to focus attention or take mitigating actions. The development and maintenance of the SRS CA is the responsibility of the SRS Management and Operations contractor, Savannah River Nuclear Solutions, LLC (SRNS). [SRNL-STI-2009-00512]

The purpose of the Liquid Waste (LW) PA Maintenance Program is to confirm the continued adequacy of LW PAs and to increase confidence in the results of the LW PAs. The elements of the LW PA Maintenance Program are:

• Testing and applied research • Monitoring • Unreviewed Waste Management Questions (UWMQs)/Performance Assessment Review

Committee (PARC) • Special Analyses (SAs) • PA revisions

This program Implementation Plan is prepared and updated annually and submitted to the DOE Savannah River Operations Office (DOE-SR). The preparation and execution of the plan is consistent with the Disposal Authorization Statement and Tank Closure Documentation (DOE-STD-5002-2017). Beginning with the FY2010 Implementation Plan (SRR-CWDA-2010-00015), the LW PA Maintenance Program activities for the SRS LW Facilities have been contained in a separate implementation plan from that for the E-Area Low-Level Waste Facility and the SRS CA.


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The purpose for this change is to better align the documents with the current SRS contract structure. A summary of LW maintenance activities is contained in Appendix A of this report; maintenance activities for the individual PAs are summarized in Tables A.1-1 and A.1-2. Table A.1-3 further contains a roll-up cost estimate summary of all LW PA Maintenance activities. This document reflects the PA-related activities completed last fiscal year (FY) or earlier in some cases, planned for the current FY, and captured out-year activities for estimation and planning purposes. Actual work performed in the out-years will be adjusted based on new program information and will be dependent on the contract baseline funding and associated actual allocated budget for that year. Section 2.0 includes a summary of the LW PA Maintenance Program activities for the SDF and Section 3.0 contains the activities for FTF and HTF. Each section includes activities relating to the following areas:

• Annual maintenance program activities • PA development/revisions (in-progress and future) • Testing and applied research activities

Saltstone Disposal Facility Overview

As shown in Figure 1.0-1, the development and review of the SDF disposal documentation has been ongoing for many years. The Section 3116 Determination for Salt Waste Disposal at the Savannah River Site, and the supporting Basis Document, were issued by DOE in January 2006. [DOE_01-17-2006, DOE-WD-2005-001] Issuance of these documents was supported by a Vault 4 SA and the SDF Performance Objective Demonstration Document. [WSRC-TR-2005-00074, CBU-PIT-2005-00146] The current SDF PA, Revision 0, which introduced a new 150-foot diameter Saltstone Disposal Unit (SDU) design, was issued in November 2009. [SRR-CWDA-2009-00017] Issuance of the current revision to the DOE Disposal Authorization Statement (DAS) for SDF, along with DOE approval of the SDF PA, occurred in May 2012. [WDPD-12-49] Three SAs utilizing new technical information, including another SDU design revision, have been issued. The first SA (the FY2013 SDF SA) was issued in October 2013 and approved in December 2013. [SRR-CWDA-2013-00062, WDPD-14-08] The second SA (the FY2014 SDF SA) was issued in September 2014 and approved in November 2014. [SRR-CWDA-2014-00006] The FY2014 SA evaluated the performance of the current 375-foot diameter SDU design. The FY2016 SDF SA was issued in October 2016 and approved by DOE in November 2016. [SRR-CWDA-2016-00072, WDPD-17-05] The FY2016 SDF SA (SRR-CWDA-2016-00072) re-evaluated the performance of the current 375-foot diameter SDU design to reflect observed field conditions (SDU 6) and to incorporate lessons learned (SDUs 7 through 12). Figure 1.0-1 also provides an overview of primary documentation issued in support of the U.S. Nuclear Regulatory Commission (NRC) consultation and monitoring role for SDF.


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Figure 1.0-1: Disposal Document Activities for Saltstone Disposal Facility


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F-Area and H-Area Tank Farms Overview

As shown in Figure 1.0-2, Tank Farm closure document development has also been ongoing for several years.

The current FTF PA, Revision 1, was issued in March 2010. [SRS-REG-2007-00002] Issuance of the Section 3116 Determination for Closure of F-Tank Farm at the Savannah River Site (DOE-WD-2012-001) and supporting Basis Document for FTF occurred in March 2012. [DOE/SRS-WD-2012-001] In March 2012, DOE also approved the FTF Tier 1 Closure Plan including its referenced FTF PA, Revision 1. [DOE_03-28-2012, SRR-CWDA-2010-00147] Along with approval of the FTF Tier 1 Closure Plan, DOE approved the Tanks 18 and 19 Tier 2 Closure Plan, including the SA specific to Tanks 18 and 19. [WDPD-12-39, SRR-CWDA-2011-00015, SRR-CWDA-2010-00124] An SA for Tanks 5 and 6 was prepared in support of operational closure of Tanks 5 and 6 and Revision 1 of the SA was issued in January 2013. [SRR-CWDA-2012-00106] DOE approved the Tanks 5 and 6 Tier 2 Closure Plan, including the Tanks 5 and 6 SA, in May 2013. [WDPD-13-56, SRR-CWDA-2013-00014] The current HTF PA, Revision 1, was issued in November 2012. [SRR-CWDA-2010-00128] Issuance of the Section 3116 Determination for Closure of H-Tank Farm at the Savannah River Site (DOE-WD-2014-001) and supporting Basis Document for HTF occurred in December 2014. [DOE/SRS-WD-2014-001] In December 2014, following issuance of the HTF Section 3116 Waste Determination, DOE approved the HTF Tier 1 Closure Plan including its referenced HTF PA, Revision 1. [DOE-OS-2015-04-27-01, SRR-CWDA-2014-00040] An SA for Tank 16 was prepared in support of operational closure of Tank 16 and Revision 1 of the SA was issued in February 2015. [SRR-CWDA-2014-00106] DOE approved the Tank 16 Tier 2 Closure Plan, including the Tank 16 SA, in May 2015. [WDPD-15-42, SRR-CWDA-2015-00009] An SA for Tank 12 was prepared in support of operational closure of Tank 12 and Revision 0 of the SA was issued in August 2015. [SRR-CWDA-2015-00073] DOE approved the Tank 12 Tier 2 Closure Plan, including the Tank 12 SA, in September 2015. [SRR-CWDA-2015-00119, WDPD-16-17] An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016 (SRR-CWDA-2016-00078), and received DOE approval in June 2017. [WDPD-17-34] This SA updated the radiological and chemical inventories for the Type I and Type II tanks in the HTF (incorporating lessons learned from the final waste tank characterization results to date) and includes an extensive series of sensitivity analyses that provide additional information that can inform decisions regarding HTF Type I and Type II tank closure operations. Figure 1.0-2 also provides an overview of primary documentation issued in support of the NRC consultation and monitoring role for FTF and HTF under NDAA Section 3116.


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Figure 1.0-2: Closure Document Development Activities for the Tank Farms


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2.0 SALTSTONE DISPOSAL FACILITY

2.1 Saltstone Disposal Facility Performance Assessment Annual Maintenance Activities DOE M 435.1-1 requires the ongoing maintenance of all PAs. This maintenance includes a series of activities that must be performed on an annual basis. This section describes the activities required every year in support of the SDF PA. All cost estimates are assumed bounding so no increases due to escalation are applied.

2.1.1 Maintain Saltstone Disposal Facility Performance Assessment Control through Unreviewed Waste Management Question Process

Description: A formal system to evaluate disposal practice changes and proposed actions is in place for the SRS LW Facilities and is known as the UWMQ process. For SDF, the UWMQ process consists of providing UWMQ Evaluations (UWMQEs) of proposed activities or new information to ensure that the assumptions, results, and conclusions of the current PA, any current SAs, the 3116 Waste Determination, and the CA remain valid and the changes are within the bounds of the DAS.

If identified through the UWMQ process that a proposed activity or new information is outside the bounds of the current analyses, new SAs are prepared to update the technical baseline. UWMQEs and SAs will continue to be required throughout the life of the facility. For planning purposes, the estimated cost assumes that 12 UWMQEs will be prepared in FY2018 (assumptions remaining at 12 for each out-year). The estimated cost does not reflect the cost of any emergent SDF SAs. Any planned SAs for SDF are captured in Section 2.2.2.

Deliverable: Provide UWMQEs and UWMQ procedure support, as needed to support SDF operations

Expected Completion Date: Ongoing

Responsibility: SRR Waste Disposal Authority (WDA)

Estimated Cost: FY2018 through FY2022 $85K/yr

2.1.2 Prepare Annual Performance Assessment Maintenance Program Implementation Plan

Description: The purpose of the LW PA Maintenance Program is to support the continued adequacy of the current PA and SAs and to increase confidence in the results. Every year the annual LW PA Maintenance Program FY Implementation Plan is prepared and provided to DOE. Plan preparation will include review of outstanding PA and SA comments and recommendations (noted in Sections 2.2.1 and 2.2.2). The Implementation Plan outlines planned work for each FY. The cost of preparing the Implementation Plan will be shared between SDF, FTF, and HTF. See the activities described in Section 3.1.2 for FTF and HTF.

Deliverable: Issue a FY LW PA Maintenance Program Implementation Plan

Expected Completion Date: 1Q-2QFY (issued annually)

Responsibility: SRR WDA



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2.1.3 Provide General Technical Support on Saltstone Disposal Facility Performance Assessment Issues

Description: This task is to provide general technical and programmatic support on SDF PA and SA issues, NRC 3116 monitoring activities (MAs), and other regulatory issues that affect SDF operations. Activities include testing and research activity support, general project support, supporting NRC on-site observation visits and technical reviews, responding to NRC requests for additional information (RAIs), and development of resolution path forward for NRC open items. Research activity support includes monitoring of research done by outside agencies (e.g., academic research, work conducted for Hanford) as well as research performed on-site (e.g., Savannah River National Laboratory [SRNL], Savannah River Ecology Laboratory [SREL]). These activities also include support on interactions with South Carolina Department of Health and Environmental Control (SCDHEC), SRS Citizens Advisory Board (CAB), the Low-Level Waste Disposal Facility Federal Review Group (LFRG), National Academy of Sciences, and other regulatory and stakeholder bodies.

Deliverable: Provide ongoing technical support on regulatory and policy issues/activities affecting SDF operations




2.1.4 Develop and Maintain Performance Assessment Model Archive and Revision Control

Description: This task established software and hardware resources for archiving development and final PA modeling files to a read-only storage medium. In FY2014, capital infrastructure improvements were enacted on the site network, allowing for faster communication between SRNL’s high performance computing network and SRR WDA servers. This improvement increased the rate for file transfers between the two systems. SDF modeling files (for both PORFLOW and GoldSim) were copied to electronic storage devices. The storage devices are maintained onsite by SRR WDA, within a cipher-locked facility. The properties of the electronic files were set to read-only. Copies of files can be provided upon request. As needed, additional storage devices will be purchased to provide sufficient disk space for maintaining a record of all related model files.

Deliverable: Establish process (completed in FY2014) and maintain after implementation




2.1.5 Conduct Annual Saltstone Disposal Facility Performance Assessment Validation Description: The purpose of the LW PA Maintenance Program is to confirm the continued adequacy of the 2009 SDF PA and to increase confidence in the results of that PA. A requirement of the maintenance program is to conduct an annual review of the disposal facility


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activities. The annual PA review is conducted in a systematic manner that incorporates the following considerations:

1. Radionuclide inventories, waste volumes, and waste types disposed throughout the year 2. Testing and research activities performed during the year 3. Results of PA monitoring conducted in accordance with the PA Monitoring Plan for

the SDF

The above factors are reviewed annually to confirm the adequacy of the current facility PA, and to evaluate the need to conduct SAs or prepare a revision to the PA. The results of the review are documented in an annual review report for the current SDF PA.

Deliverable: Issue a FY PA annual review report

Expected Completion Date: 1-2QFY (issued annually)



2.1.6 Maintain Saltstone Disposal Facility Performance Assessment Closure Plan Description: The closure plan that complies with DOE M 435.1-1 for SDF must be maintained and modified as needed to reflect facility changes. The SDF closure plan is reviewed annually to determine if a revision is required. Revision 1 to the SDF Closure Plan was completed in September 2015. [SRR-CWDA-2013-00037] The revision incorporates design changes to the SDF reflected in the FY2014 SDF SA, including the incorporation of the 375-foot diameter SDU design. No revisions to the SDF Closure Plan were issued in FY2017.

Deliverable: Review closure plan annually and revise as necessary

Expected Completion Date: Reviewed annually



2.1.7 Maintain Saltstone Disposal Facility Performance Assessment Monitoring Plan Description: The SDF PA Monitoring Plan that complies with DOE M 435.1-1 must be maintained and modified as needed to reflect facility changes. The Monitoring Plan is reviewed annually to determine if a revision is required. Revision 1 to the SDF PA Monitoring Plan was completed in August 2015. [SRR-CWDA-2013-00026] The revision incorporated and integrated the ongoing activities relative to new ground water wells and the 375-foot diameter SDU design at the SDF. No revisions to the SDF PA Monitoring Plan were issued in FY2017.

Deliverable: Review Monitoring Plan annually and revise as necessary

Expected Completion Date: Review annually




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2.2 Saltstone Disposal Facility Performance Assessment Development/Revisions The 2009 SDF PA provides the technical basis and results to be used in subsequent documents to demonstrate compliance with pertinent requirements of DOE M 435.1-1 and associated references and Title 10 Code of Federal Regulations (CFR) Part 61, Subpart C as required by NDAA Section 3116. The 2009 SDF PA incorporates the existing SDUs (SDUs 1 and 4) and anticipated future SDUs at that time. Approval of the 2009 SDF PA and issuance of an updated DAS was received in 2012. [WSRC-RP-92-1360, SRR-CWDA-2009-00017, WDPD-12-49]

2.2.1 Prepare Out-Year Saltstone Disposal Facility Performance Assessment Revisions

Description: The 2009 SDF PA has been issued and implemented and work has begun to revise the 2009 SDF PA. WDA performed an SA in FY2016 as discussed in Section 2.2.2. While the work has begun on a PA revision, the completion date will be dependent on the receipt of Research and Development results and adequate funding. The future SDF PA revision will include the following items at a minimum:

• Analyses and results contained in all SAs that have been completed to date; • Analyses and results of all UWMQEs completed to date; • Consideration of new information generated through applied research; • Changes to and/or optimization of disposal unit design; • Updates to General Separations Area Database and calibration for Z Area; • Changes in site future land use plans or closure plans; and • Changes to PA guidance documents requirements.

In addition, the future SDF PA revision will include the following items resulting from applied research:

• DLM results; • Updates to closure cap infiltration rates: and • Updates to cementitious material degradation rates.

Deliverable: Issue PA revision

Expected Completion Date: FY2019


Estimated Cost: FY2018 $500K, FY2019 $350K, FY2020 through FY2022 $0

2.2.2 Saltstone Disposal Facility Special Analyses WDA performed a SA in FY2016 (SRR-CWDA-2016-00072) that adjusted the physical locations of the SDUs, re-evaluated the performance of the current 375-foot diameter SDU design to reflect observed field conditions (SDU 6) and to incorporate lessons learned (SDUs 7 through 12). The results presented in the 2009 SDF PA or FY2014 SA were not significantly impacted by the new information presented in this FY2016 SDF SA. Currently, there are no SDF SAs scheduled for the period covered by this LW PA Maintenance Program. Development and documentation of responses to any potential NRC RAIs on the FY2016 SA is covered under Section 2.1.3.


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Deliverable: No additional SAs scheduled

Expected Completion Date: Not Applicable


Estimated Cost: FY2018 through FY2022 $0/yr

2.3 Saltstone Disposal Facility Performance Assessment Testing & Research Activities This section contains the PA-related testing and research activities that are being performed as (1) part of the ongoing maintenance activities aimed at reducing uncertainty in the SDF PA and SA models, or (2) verification sampling and analysis of material properties used in the PA (i.e., verification of emplaced saltstone properties and properties of saltstone cured to actual vault temperature and humidity). As ongoing research provides new information or reduces uncertainty, this information will be evaluated (via the UWMQ or SA process described in Section 2.1.1) against the information used as a basis for modeling.

After saltstone production operations have ceased, a closure cap will be installed over the SDF to mitigate the infiltration of water through the disposal units and the saltstone waste form. There are key questions related to closure cap design and performance that could affect the results of the modeling (e.g., plugging of the drainage layer). In addition, the SDF PA and SAs suggest that parameters most sensitive to SDF performance are related to the saltstone waste form. [SRR-CWDA-2009-00017, SRR-CWDA-2013-00062, SRR-CWDA-2014-00006, SRR-CWDA-2016-00072]

As such, in the near term, resources were prioritized to support testing and modeling research activities related to key parameters of the saltstone waste form and the disposal units as reflected in Figure 2.3-1. Research was also initiated in FY2017 to re-evaluate the closure cap infiltration rates.

Figures 2.3-2, 2.3-3, and 2.3-4 provide simple illustrations of some of the parallel testing efforts described in detail in this section. From an overall perspective, the illustrations depict how testing and research activities and ongoing testing of the saltstone waste form are being selected using an integrated, systematic approach.

Funding estimates have been made for each ongoing or anticipated activity. While actual work performed is always dependent on current funding and priorities, the tables in Appendix A provide a general idea of the work expected to be performed over the next several years.

In September 2013, the NRC revised their NDAA Section 3116 monitoring plan for SDF. This NRC Monitoring Plan provides expectations for closing TER issues. Note that the monitoring factors in the NRC Monitoring Plan cover both the NRC concerns from the 2012 NRC TER and


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Figure 2.3-1: Saltstone Research, Development, and Testing Program Elements


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the previous Open Issues that had not been closed from an earlier version (Revision 0) of their NRC Monitoring Plan. [ML13100A113] Appendix B of this FY2018 Implementation Plan provides a summary of monitoring factors in the context of testing and research activities related to the SDF.

2.3.1 Critical Property Testing Several parameters are essential to reducing uncertainty in PA values, specifically in the areas of hydraulic conductivity, technetium and iodine desorption, and distribution coefficients (Kds) for cured saltstone. The current strategy aims at reducing uncertainty in critical property parameters, including current research efforts as shown in Figure 2.3-2.

The maintenance activities presented in this section concern critical values such as Kd values, hydraulic conductivities, technetium and iodine desorption, and water retention characteristics. The transport modeling relies on such values to make informed predictions about system behaviors over long periods. It is therefore desirable to reduce uncertainty in these parameters where possible. The maintenance activities presented here are intended to reduce uncertainty around properties in the modeling.

There were many Kd values used during the transport modeling. Values were used for greater than 40 radionuclide species (note that radioisotopes of the same element have the same Kd and solubility values) and eight solid phases. Additionally, ranges for each value and their type of statistical distribution (e.g., normal or log normal) provide additional confidence in the results. These values were generated in different ways depending on the availability of experimental data. Part of these efforts focuses on reducing uncertainty in the Kd values that are most important to the transport modeling.

Several studies from FY2011 and FY2012 have influenced the direction of continued research on properties considered critical to the performance of saltstone. A study on the hydraulic and physical properties of saltstone and their correlation to the mix and curing conditions has prompted further investigation into the effects of cure temperature and curing conditions on saltstone performance properties. [SRNL-STI-2011-00665, SRNL-STI-2012-00558] Studies on the technetium (Tc) Kd under anaerobic conditions began in FY2011 and continued in FY2012. The sorption experiment started in FY2011 and continued until equilibrium was reached. [SRNL-STI-2011-00716, SRNL-STI-2012-00596] A column study was initiated to independently substantiate these values using an alternate method that provided additional information. [PNNL-21723]

More recently, studies conducted in FY2013 through FY2017 have predominantly focused on the following:

• Measurement of depth dependent saltstone oxidation, • Sorption properties (i.e., Kd) of saltstone contaminants in the presence of SDF soils and

SDU concrete, • Determining the influence of curing environment on the saturated hydraulic conductivity

(SHC), • Technetium leaching characterization from intact (not size reduced) saltstone samples, and • Successfully extract and analyze saltstone samples from an actual SDU.


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Figure 2.3-2: Critical Property Testing Strategy


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In FY2013, saltstone oxidation studies used chromium (Cr) as a surrogate for technetium, but due to the inconclusiveness of the data and potential dissimilarities between chromium and technetium redox behavior, studies for FY2014 and FY2015 were directed towards characterizing the oxidation states of technetium, sulfur (S), and iron (Fe) in Tc-spiked simulant saltstone samples. Reduced iron and sulfur species contained in blast furnace slag (BFS) are anticipated to play a role in the reduction, and subsequent immobilization, of the technetium contained in saltstone. With respect to iron, although the FY2014 data indicated the presence of ferrous (Fe2+) iron in early aged samples, for longer aging times the ferrous iron was found to be completely oxidized to ferric iron (Fe3+). For sulfur, the dominant species irrespective of aging time was sulfate (SO4

2-). However, a reduced sulfur species in the form of thiosulfate was also detected. The proportion of thiosulfate was observed to increase as a function of depth but decrease as a function of time indicating progressive oxidation of reduced sulfur species. With respect to technetium speciation, the majority (approximately 70%) of the Tc(VII) originally spiked into the saltstone samples was reduced to the Tc(IV) oxidation state; however, the exact Tc-compounds present in the cured saltstone could not be definitively identified. The proportion of Tc(VII) reduced after incorporation into the cementitious matrix is, however, particularly significant since the initial Tc-loading concentration was over two orders of magnitude greater than that measured in the Tank 50 waste solution. [SRRA042328-000004] Further speciation analysis of Tc-spiked, simulant saltstone samples is not envisaged.

Leaching studies with respect to technetium and rhenium (Re) (a non-radioactive surrogate for technetium) were conducted through FY2014 and FY2015. [SREL Doc. R-14-0007, SREL Doc. R-15-0003] Two separate leaching methods were used. Firstly, the U.S. Environmental Protection Agency (EPA) prescribed leaching technique (EPA Method 1315) in which samples are submersed in a solution and the technetium (or rhenium) concentrations in the solution measured at prescribed times between 0 and 63 days. [EPA_Method_1315] The effective diffusivities determined via EPA testing for technetium and rhenium are approximately 3E-10 and 3E-08 cm2/s, respectively. Given the diffusivity of rhenium is two orders of magnitude greater than technetium this may indicate that rhenium is not an appropriate surrogate for technetium when evaluating leaching behavior. The second leaching method utilizes a permeameter (typically used for measuring hydraulic conductivity) to force solution through a monolithic sample; the effluent emerging from the sample is subsequently characterized for technetium (or rhenium) content. This method, termed the Dynamic Leaching Method (DLM), was specifically developed to inform the PA model since it is anticipated that the transport of a known volume of solution through the sample can be directly equated to pore volume exchange in the actual saltstone monolith. This method continues to be used and findings are provided in the latter text. Note that actual saltstone cores retrieved from SDU Cell 2A have also been analyzed with respect to the two aforementioned leaching tests; the data derived from these samples is discussed in Sections 2.3.1.2 and 2.3.3.

Though several studies from previous years measured the sorption properties of saltstone contaminants, the studies were not performed in the presence of actual SDF soils. In FY2013, a study was performed to measure sorption properties in the presence of SDF soils. Virtually all contaminants exhibited similar sorption properties to those observed in previous studies.


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[SREL Doc. R-13-0004, SREL Doc. R-13-0005] In FY2014, an attempt was made to elucidate more quantitative iodide Kd values since the FY2013 measurements had only determined that the values in the presence of different soils and solution chemistries were less than 1. The desired improvements were not realized for the FY2014 experiments, and these activities were discontinued. [SREL Doc. R-14-0005] Based on the results of the FY2014 testing, additional testing was planned for FY2016 utilizing radioiodine in order to distinguish between natural iodine that is present in the saltstone dry feed materials or SRS soils, and the intended I-spike added to establish I Kd. In the previous work a non-radioactive I-spike was used that could not be distinguished from iodine contained in the saltstone dry feeds and soils. In FY2017 Kd measurements were conducted with saltstone samples spiked with I-129 in the presence of SDF soils and SDU concrete. These experiments continued to indicate I-129 Kd values less than 1.

With respect to the effects of curing environment on the SHC of saltstone, simulant saltstone samples were cured using a temperature profile (with a maximum temperature of 65 oC) recorded in an SDU at relative humidities (RHs) of 40%, 90%, and 100%. As anticipated, the lowest hydraulic conductivities were recorded for those samples cured in a water-saturated environment. [VSL-13R3010-1] Additional samples were also cured according to thermal profiles that qualitatively mimicked the array of profiles recorded in an SDU. Three thermal profiles were considered with each profile starting at room temperature and subsequently ramping at 10 °C per day to peak temperatures of 35, 45, and 55 °C. After the peak temperature was attained the samples were cured for 90 days. Saturated hydraulic conductivity was measured after 28- and 90-day curing durations. Though some differences were apparent in the saturated hydraulic conductivities measured for the 28-day samples the values essentially normalized after 90 days, irrespective of the peak curing temperature. Saturated hydraulic conductivities ranged from 1.3E-09 to 2.4E-09 cm/sec. [VSL-14R3210-1] No additional work is currently planned in this area, additional details outlining the results of the studies was provided in Section 2.3.1.3 of the FY2015 Implementation Plan. [SRR-CWDA-2014-00108]

Considerable effort was directed towards developing a strategy for retrieving SDU-emplaced saltstone and subsequently correlating properties, such as SHC, with samples retrieved from the Saltstone Production Facility (SPF) processing room, and laboratory-processed saltstone. Retrieval of fresh grout samples from the SPF and development of a mock-up to assess core drilling as a method to ultimately retrieve emplaced grout from the SDU was performed in FY2013. [SRR-CWDA-2014-00059] For FY2014 both the core-drilling and sample extraction techniques were further assessed and optimized, and drilling of SDU Cell 2A was completed in FY2015. In April and May of 2015 SRR extracted approximately 191 inches of 2-inch diameter core material from SDU Cell 2A. The core material was inerted as soon as possible upon extraction from SDU Cell 2A and transported to SRNL. The core material was stored in an anaerobic chamber in SRNL. [SRR-CWDA-2015-00066] Analyses conducted in FY2016 on the saltstone cores included SHC, Kds for various radionuclides, measure of acidity or alkalinity of a solution (pH), measure of reduction (or oxidation) potential (Eh), Tc(VII) to Tc(total) ratio, and density, porosity, and moisture content. Additional detail on the analysis of the SDU Cell 2A core material is provided in Section 2.3.3 of this FY2018 Implementation Plan.


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2.3.1.1 Measurement of Distribution Coefficients in SRS Subsurface Sediments Description: This study will measure the Kds for various species under oxidizing and reducing conditions in actual subsurface sediments retrieved at SRS (actual SDF soils used). Additionally, testing will measure the impact of high pH cement leachate on the subsurface sediments Kd. Expected Benefit: To reduce the uncertainty in the SDF, FTF, and HTF PAs with respect to the Kd values assumed for radionuclides in soils.

FY2013: Batch partitioning experiments using multi-element spike solutions were conducted to determine the impact of cementitious material leachate (across a pH range of 4.8 to 12.2 under oxic and anoxic environments) on the Kd for a disparate range of 17 contaminants and non-radioactive chemical analogs. [SREL Doc. R-13-0004] In addition to measuring the impact of cement leachate on contaminant partitioning, this study also sought to establish the effects of using site-specific soil materials collected at the SDF. The adsorption characteristics of the following contaminants were tested, and subsequently characterized, using Inductively Coupled Plasma – Mass Spectrometry (ICP-MS): chromium (Cr), nickel (Ni), arsenic (As), selenium (Se), strontium (Sr), cadmium (Cd), iodine (I), antimony (Sb), cesium (Cs), barium (Ba), lanthanum (La), cerium (Ce), europium (Eu), rhenium (Re), thallium (Tl), thorium (Th), and uranium (U).

Most Kd values established via this study were equivalent to, or higher than, previously reported values that are currently used in the 2009 SDF PA. Notable exceptions included arsenic, antimony, and selenium. Arsenic, antimony, and selenium had cement leachate impacted Kd values that were much lower than the literature values reported in Geochemical Data Package for Performance Assessment Calculations Related to the Savannah River Site (SRNL-STI-2009-00473), which served as the basis for Kd values used in the 2009 SDF PA. Specifically, in this recent testing, arsenic was measured to be 13 to 14 mL/g, antimony was 5 to 7 mL/g, and selenium was 17 to 19 mL/g; whereas the previous values were estimated to be 140 mL/g, 3,500 mL/g, and 1,400 mL/g, respectively. However, note that the Kd value for selenium under natural conditions was higher than the current modeling value (i.e., greater than 1,400 mL/g). The FY2014 SDF SA included an analysis (Section 5.6.6.4.3) that demonstrated that applying the lower cement leachate impacted Kd values for selenium would not have a significant impact to doses within the 10,000-year period of performance. [SRR-CWDA-2014-00006]

Niobium (Nb) and radium (Ra) were analyzed as single element spike solutions using the same leachate and environmental variables considered for the multi-element contaminant solutions. [SREL Doc. R-13-0005] Niobium Kd values for the SDF soil were high (Kd greater than 1,200 mL/g) for all treatment solution chemistries, regardless of redox status. Radium Kd values were high (Kd greater than 1,500 mL/g) for both cementitious materials (i.e., reducing saltstone and conventional cement), regardless of background solution chemistry or redox status. This value is considerably higher than previous modeled values (approximately 100 mL/g) for relatively young oxidizing and reducing cement materials and weathered/aged oxidizing and reducing cement materials (approximately 70 mL/g).

FY2014: The Kd values observed for iodide in the FY2013 study (less than 1.0 mL/g) were consistent with previously reported values regardless of leachate pH and redox conditions.


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Since iodide is a major dose driver with respect to the saltstone modeling, absolute Kd values are preferred as inputs. However, for previous studies absolute values for iodide Kd have been difficult to measure since limited iodide adsorption makes it difficult to quantify major changes in iodide partitioning. FY2014 work considered modification of the test conditions to focus on improved quantification of low-level iodide sorption; in particular, a high sorbent-to-solution ratio (i.e., 15 g sorbent to 30 mL test solution) and a low iodide test concentration (1.0E-01 mg/L) were used in an effort to further refine the low Kd values. Iodide partitioning was evaluated in the presence of SDF soil materials and an SDU concrete sample, and under four different background solution chemistries chosen to mimic various stages of saltstone/cement aging throughout the lifetime of the closed facility, from initial cementitious material formation and curing through weathering and eventual far-field leaching in the underlying soil environment. Two alkaline treatment solutions (pH between 11 and 12), termed portlandite and salt waste leachate solution, were chosen to mimic the initial stages of cement weathering where leachate composition has the potential to dramatically alter contaminant partitioning to SRS soils and sediments. The remaining two treatment solutions, termed artificial groundwater (AGW) and calcite-saturated solution, were used to approximate conditions for contaminant partitioning after extensive weathering of the cementitious materials and/or dilution with non-impacted SRS groundwater. All partitioning experiments were conducted under oxic and anoxic (2% H2 atmosphere) conditions to evaluate the overall impact of redox status on contaminant partitioning to the SDF soil and SDU concrete. While some iodide sorption was observed for the SDF soils under acidic pH conditions, iodide

partitioning levels were quite low for the soil under alkaline conditions, making it difficult to discriminate between partitioning to the solid-phase from experimental variation and analytical error. Partitioning to the SDU concrete was low regardless of the treatment solution due to the inherently high pH of the final concrete test solutions.

FY2016: Based on the results of the FY2014 testing, additional testing was performed in FY2016 utilizing radioiodine in order to distinguish between natural iodine that is present in the saltstone dry feed materials or SRS soils, and the intended I-spike added to establish iodine Kd. In the previous work a non-radioactive I-spike was used that could not be distinguished from iodine contained in the saltstone dry feeds and soils.

The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was revised in July 2016 (previous revision issued in March 2010). This update incorporated the numerous experiments and geochemical measurements have been conducted since 2010 (including those discussed above), resulting in new recommended input values for modeling. The revised geochemical data package integrates recent documented geochemical results, including radionuclide Kd values, solubility values, and cementitious impact factors, and includes a critical evaluation of these values with respect to existing values to assess potential impacts. The primary changes included in the latest revision are as follows:

• Included results from more than 70 new studies, including 22 new SRS studies providing site specific information related to SRS PAs.

• The impact of cellulose-degradation product on Kd values was dropped from the data package because it was determined not to be needed (SRNL-STI-2012-00138).


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• In the past, the number of significant figures for Kd and SHC values were not held constant. Only one significant figure is used in this data package to better represent the accuracy of these parameters.

For the vast majority of the dataset, the new information supported the use of the previous Kd inputs. Perhaps the most significant deviations from the 2010 dataset were decreased Kd values for I-129 in the presence of reducing cement. In some cases (environmental Stages I and III) the Kd decreased from 4-5 to zero essentially indicating no retention effect of the reducing cement for I-129. For environmental Stage II the I-129 Kd decreased from 9 to 2 again indicating decreased retention capability of reducing cement for radioiodine. In contrast, newly established Kd values for Pu, Ra, Sr, and U were shown to be consistent with previous inputs or increased in comparison to the previously utilized inputs. For environmental Stages I and II in the presence of reducing cement, Tc-99 Kd decreased from 5,000 to 1,000.

FY2017: In FY2017, the investigation of I-129 in the presence of SRS soils and cementitious materials was conducted. [SREL Doc. R-17-0004] While some I-129 sorption was observed in the presence of SDF soils under acidic pH conditions (Kd ≈ 0.18 to 1.0 mL/g), I-129 partitioning levels were quite low for the soil under alkaline conditions (<0.18 mL/g), making it difficult to discriminate between partitioning to the solid-phase from experimental variation and analytical error. For the SDU concrete samples, the equilibrium pH was quite high regardless of the initial background solution, indicating that the concrete had a large impact on solution chemistry. In fact, all of the observed Kd values for the SDU concrete materials were < 0.18 mL/g.

FY2018: For FY2018, the retention of I-129 will be evaluated with respect to intact saltstone monoliths using DLM since this sample format is more representative of the field configuration in comparison to size-reduced (crushed) samples utilized for partitioning experiments. Any future work will be described under Section 2.3.3 of this FY2018 Implementation Plan

Deliverables: None

Expected Completion Date: Task complete



2.3.1.2 Contaminant Leaching Characteristics from Saltstone Monolith Description: The purpose of this study is two-fold. The first element of the investigation is to evaluate test methods with respect to saltstone leaching utilizing rhenium and stable iodine (I-127) as non-radioactive surrogates for Tc-99. Test methods to be employed include a standardized semi-dynamic leaching test, EPA Method 1315, Mass Transfer Rates of Constituents in Monolithic or Compacted Granular Materials Using a Semi-Dynamic Tank Leaching Procedure, and a dynamic leaching test that will be developed as part of this scope. [EPA_Method_1315] The dynamic leaching methodology will use the flexible-wall permeameter apparatus that is more commonly used for measuring the SHC of saltstone. The intent is to force leachate through the interior of the saltstone monolith to mimic the eventual ingress of water into saltstone and subsequent pore volume exchange to establish the dynamic leaching behavior of saltstone contaminants. The second element of the investigation is to


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utilize the tests methods (optimized with non-radioactive surrogates) for characterizing the leaching behavior of saltstone samples spiked with Tc-99 and I-129 in addition to the characterization of saltstone cores retrieved from SDU Cell 2A.

Expected Benefit: This task will provide empirical leaching (diffusion and solubility) data for Tc-99, I-129, and potentially other saltstone contaminants that can be used as direct inputs to the saltstone transport models. In addition, the development of a dynamic leaching test will provide new information regarding the leaching of saltstone associated with multiple pore volume exchanges.

FY2014: Contaminant mass transfer for the I- and Re- spiked monoliths was evaluated using EPA Method 1315. [EPA_Method_1315] The mass transfer test was conducted using an AGW simulant as the leachate in equilibrium with three different test atmospheres: (1) ambient laboratory atmosphere (oxic), (2) ultra-high purity N2-purged atmosphere (anoxic), and (3) 98% N2 and 2% H2 atmosphere (anoxic reducing). Similar leaching patterns were observed for rhenium and iodine in monoliths cured for 30 and 90 days, with very little difference observed for a given test analogue in the three atmospheric test environments. Maximum eluent concentrations of approximately 1.5E-02 and 9.3E-03 mg/L were observed (after 28 days) for rhenium and iodine, respectively. The cumulative release history for both rhenium and iodine conformed to a diffusion-controlled mechanism. For rhenium, the lack of response to extraction atmosphere may reflect the difficulty in achieving reduction, with similar amounts of Re(VII) available for release regardless of the atmospheric conditions.

The dynamic leaching method was based on American Society for Testing and Materials (ASTM) D5084, which is typically used for determining the SHC of saltstone using a flexible-wall permeameter. [ASTM D5084-10] The permeameter system was used to develop the required hydraulic gradient to force the leaching solution through the sample. Cured saltstone monoliths spiked with non-radioactive iodide and perrhenate (ReO4

-) were used to evaluate the leaching method in lieu of radioactive Tc-99. Chemical analysis of the effluent leachates focused on rhenium rather than iodine because of the limited sample volumes and difficulty in stabilizing iodine for analysis by ICP-MS. Initially high levels of rhenium (2.5E-01 to 8.0E-01 mg/L) were observed in the monolith leachates, which decreased to a more stable level of 1.0E-01 to 1.5E-01 mg/L with continued leaching. Initial testing utilized 2-inch diameter monoliths that required substantial hydraulic gradients (approximately 20 psi) for consistent flow. During these tests, the SHC of the saltstone monolith tended to decrease over time. At least a portion of the reduction in hydraulic conductivity may be attributed to the formation of gas bubbles, presumably the result of degassing associated with the drop-in pressure along the column. Furthermore, the caustic nature of the effluent and the formation of gas bubbles also confounded the use of flow-through electrodes for continuously monitoring effluent chemistry (i.e., pH and dissolved O2). Far less degassing was observed in subsequent column experiments conducted at lower hydraulic gradients (10 psi), but with a proportional reduction in column leaching rate. [SREL Doc. R-14-0006, SREL Doc. R-14-0007]

FY2015: The standardized semi-dynamic leaching, EPA Method 1315, was used to evaluate the leaching characteristics of Tc-spiked saltstone samples cured for three and six months. [EPA_Method_1315] A Tc-spiking concentration of 1.2E-02 mMol/L was utilized for all samples, which is consistent with technetium levels detected in Tank 50 waste solution.


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Similar to FY2014, tests were conducted with an AGW solution equilibrated with respect to three different atmospheres (oxic, anoxic, and reducing). The leaching rate of Tc-99 was observed to decrease over the course of testing but exhibited no clear response to either curing duration (3 or 6 months) or test atmosphere (oxic, anoxic, and reducing). In comparison to Re-spiked samples, the proportion of technetium leached during testing was significantly less than rhenium (i.e., less than 0.75% for technetium compared to 6 to 9% for rhenium). Rhenium is typically considered as a non-radioactive surrogate for technetium but this data, and known differences in redox potential between the two elements, imply that rhenium does not exhibit equivalent leaching behavior to technetium and as such may not be a legitimate surrogate. The effective diffusivities determined via EPA testing for technetium and rhenium were 2.0-3.7E-10 and 2.7-3.5E-08 cm2/s, respectively.

Dynamic leaching tests utilizing the method developed using Re-spiked samples in FY2014 were conducted on Tc-spiked samples in FY2015. Based on the occurrence of effluent degassing in FY2014, which appeared to prevent the acquisition of consistent data, a bladder accumulator was incorporated into the experimental set up to restrict contact between the “test” solution saturating the column and the solution in the permeameter. This modification was successful in excluding dissolved gases from the pressurized inlet solution, and subsequent bubble formation in the sample and in the effluent solution. Similar to the EPA measurements, technetium and rhenium were spiked at levels equivalent to those measured in Tank 50 waste. After collecting approximately 45 to 50 mL of leachate for each of the Tc- and Re-spiked samples, the cumulative proportions of technetium and rhenium (as a percentage of the original spike amounts) leached were 0.3% and 10%, respectively. The disparity in the data again suggests that rhenium may not be a suitable non-radioactive surrogate for technetium at least in regards to leaching behavior. Based on sample size and a measured porosity of 60%, 45 to 50 mL equates to the exchange of approximately 1 to 2 pore volumes though it is acknowledged that not all of the porosity may be involved in liquid transport through the sample. [SREL Doc. R-15-0003]

FY2016: Both EPA Method 1315 and dynamic leaching testing were continued in FY2016 and encompassed evaluation of radionuclide-spiked saltstone simulants and actual saltstone cores extracted from SDU Cell 2A. The data from these studies is provided in SREL Doc. R-16-0003.

With respect to EPA method testing, Tc-99 leaching rates for the spiked saltstone samples appeared to be sensitive to curing duration and the reduction capacity of the BFS used in making the grout. Due to supply cessation of a historically utilized BFS, an alternate BFS source was sought and approved for use in processing future saltstone batches at SRS. Longer curing times and higher reduction capacity for the as-received BFS resulted in lower effective diffusivities. A Tc-spiked, simulant saltstone processed using BFS with a reduction capacity of 1,600 µeq/g and cured for 6 months indicated an effective diffusivity of 5.7E-12 cm2/sec compared to 3.0E-10 cm2/sec for a sample processed using BFS with a reduction capacity of approximately 700 µeq/g, and cured for 3 months. Reduction capacity can vary between different BFS sources due to differences in the concentrations of components, in particular sulfur and iron, which are known reductants. As anticipated leaching rates and effective diffusivities from the simulant samples for poorly sorbing contaminants like NO3

- were much


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higher than for Tc-99 (E-07 to E-08 cm2/sec for nitrate compared to E-10 to E-12 cm2/sec for Tc-99).

For DLM testing, the permeameter system was upgraded to accommodate three test samples at the same time by using the laboratory air compressor to provide the driving force for leaching. The three materials tested include a Tc-99-spiked sample described in SREL Doc. R-15-0003, and two actual radioactive saltstone cores extracted from SDU Cell 2A. Measurement of Tc-99-spiked sample has been ongoing for almost 15 months with approximately 230 mL of leachate being collected from the sample (equivalent to about 6.25 pore volumes). The concentration of Tc-99 in the leachate up to approximately 4.5 pore volumes fluctuated between 5E-09 and 1E-08 mol/L with a cumulative Tc-99 release of 0.6%. However, leachates collected after the exchange of 4.5 pore volumes indicated a rapid increase in Tc-99 concentration to around 1E-07 mol/L; at the time of writing the cumulative Tc-99 release from the simulant saltstone sample was approaching 2.5% (after the exchange of 6.25 pore volumes). In contrast, approximately 50% of the nitrate in the sample has been released which verifies the poor nitrate retention observed for the aforementioned EPA tests. It is also noteworthy that the nitrate concentration did not indicate a similar spike to Tc-99 after the exchange of approximately 4.5 pore volumes. Phenomena responsible for the sudden increase in Tc-99 leaching have not been established.

The Tc-99 concentrations and cumulative Tc-99 release for the SDU Cell 2A samples are higher than previously observed for the Tc-spiked, simulant saltstone. Tc-99 leachate concentrations for both SDU Cell 2A samples have been measured at approximately 1E-07 to 4E-07 mol/L, an order of magnitude higher than concentrations measured for the Tc-spiked sample. In contrast, the concentration of nitrate in leachates of all three samples is much more consistent; both SDU Cell 2A samples and the Tc-spiked sample indicate an initial concentration of around 1.0 mol/L that subsequently decreases gradually to between 0.1 to 0.4 mol/L.

FY2017: During the FY2017 testing of the Tc-spiked saltstone, an experimental anomaly occurred in which the sample confining pressure was lost and it is possible that the sample was exposed to the atmosphere. The Tc-99 concentration following this anomaly increased by a factor of three (from 1.2E-07 to 3.6E-07 mol/L); the anomaly is highlighted in Figure 2.3-3 which indicates the change in Tc-99 and nitrate concentration as a function of pore volumes exchanged. More significant, however, was a concurrent order of magnitude increase in the SHC from 4.0E-10 cm/sec to 4.0E-09 cm/sec. The SHC did not decrease to its former value and remains around 2.0E-09 cm/sec (Figure 2.3-4). As such there is concern that the sample, and therefore the data, after this event may be compromised. However, given the extended length of time the sample had been tested, a decision was made to continue the analysis. At this point (December 2017), the sample has been evaluated for approximately 30 months. Fifteen pore volumes have been exchanged and the Tc-99 and NO3

- concentrations are currently 3.3E-08 mol/L and 0.02 mol/L, respectively (Figure 2.3-5). The pH of the leachate as a function of pore volumes exchanged is indicated in Figure 2.3-4. In addition, the Tc concentration is plotted as a function of leachate pH in Figure 2.3-6. The thermodynamically predicted Tc phases and their solubility with respect to pH are also indicated on Figure 2.3-6. The Tc-99 concentrations and associated pH are in good agreement with the thermodynamically predicted solubility limit of reduced TcO2.xH2O phases. With respect to


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cumulative % leached from the saltstone sample the Tc-99 is 11% and the NO3- is 94% (Figure

2.3-5).

Figure 2.3-3: Leachate Concentrations for Tc-spiked DLM Sample

Figure 2.3-4: SHC and pH Plotted as a Function of Pore Volumes Exchanged for Tc-

spiked DLM Sample

1E-10

1E-09

1E-08

1E-07

1E-06

1E-05

1E-04

1E-03

1E-02

1E-01

1E+00

1E+01

0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0∑Pore Volumes Exchanged

Leac

hate

Con

c. (m

ol/L

)

NO3- (0.02 mol/L)

99Tc (3.3E-08 mol/L)

Experimental Anomaly

10

11

12

13

14

1E-12

1E-11

1E-10

1E-09

1E-08

1E-07

0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0∑Pore Volumes Exchanged

SHC

(cm

/sec

)

pH

Ksat (2.4E-09 cm/sec)

pH (10.8)

Experimental Anomaly


Page 29 of 143

Figure 2.3-5: Cumulative % Tc-99 and Nitrate Leached from Tc-spiked DLM Sample

Figure 2.3-6: Leachate pH and Tc Concentrations for Tc-spiked DLM Sample

The DLM samples described thus far have been evaluated on the originally developed DLM system in which a constant pressure is set and the flow rate allowed to vary based on changes with respect to the sample SHC. The newly designed system utilizes mechanical pumps in which a constant flow rate through the sample can be set. It should be noted that this new system has not performed as intended since the flow rates continue to vary based on changes in sample SHC. The original intent for being able to control the flow rate was two-fold. Firstly, the DLM experiments as currently designed are slow as indicated by the transport of only 15 pore volumes in 30 months for the aforementioned Tc-spiked sample. Secondly, it was

0

20

40

60

80

100

0.0 2.0 4.0 6.0 8.0 10.0 12.0 14.0 16.0

∑Pore Volumes Exchanged

NO3- (93.6%)

99Tc (10.6%)

% C

onst

ituen

t in

Lea

chat

e

1E-09

1E-08

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9 10 11 12pH

Solu

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y (m

ol/L

)

TcO2.1.6H2O

TcO2.2H2O

KTcO4

NaTcO4.4H2O


Page 30 of 143

envisaged that being able to control the flow rate would enable the impact of permeant residence time to be determined. To date, four samples have been evaluated on this revised DLM system and while the desired flow rates were set between 0.2 and 1.5 mL/day all samples are currently operating at rates at or below 0.1 mL/day. As such it may not be possible to determine residence time effects on leachate chemistry with the current experimental set up. These samples have had less than a pore volume exchanged and as such it is too early to make any determinations with respect to sample-to-sample similarities or differences. In comparison, to the aforementioned Tc-spiked samples, the Tc-99 concentrations in the leachates are high and currently measured at 1.0E-06 mol/L. However, with leachate pH values in the range of 11.8 to 12.8 these concentrations are generally aligned with the thermodynamically-predicted solubility limits for reduced TcO2.xH2O phases.

FY2018: Plans for FY2018 are predominantly focused around continued DLM evaluation of the Tc-spiked samples and the SDU Cell 2A core(s). There are also two additional areas of emphasis. The first is to analyze the leachates from the Tc-spiked samples with respect to the concentration of constituents that are relevant to the pore solution chemistry of saltstone. Key constituents of interest include Na, Ca, K, Al, Mg, Fe, Si, NO3

-, NO2-, CO3

2-, and SO42-. It is

envisaged that in addition to the influence of pH on Tc-99 solubility, changes in the pore solution constituent concentrations may also influence the Tc-99 leaching characteristics. The second is to utilize DLM to evaluate the leaching behavior of I-129 from spiked laboratory-prepared samples. I-129 is the most significant dose driver and aforementioned partitioning measurements indicate that it is poorly retained in saltstone. However, those experiments require the saltstone sample to be crushed thereby exposing a larger proportion of the saltstone sample to the leachant solution. In reality, saltstone is comprised of a tortuous, sub-micron pore network which is anticipated to impact the leaching characteristics of I-129.

Deliverable: Annual Technical Reports




2.3.2 Ongoing Studies Maintenance activities presented in this section will focus on meeting information needs relevant to various aspects of system performance. The overall strategy to gain a better understanding of ongoing studies is shown in Figure 2.3-7. Various programs discussed below include:

• Long-Term behavior of radionuclides in Lysimeters

• Cementitious materials degradation due to radiation exposure

• Closure cap performance

• Hydraulic degradation in SDU concrete


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2.3.2.1 Long-Term Radiological Lysimeter Program Description: Understanding the long-term behavior of radionuclides in saltstone is essential to models that project this behavior over thousands of years. The objective of this task is to measure the release from radioactive samples by placing cementitious materials (saltstone or grout) and soils spiked with a suite of radionuclides/analogues including cesium, plutonium (Pu), iodine, and technetium in lysimeters to be placed in an outside environment. Measurement will target solubility and Kd values in soil and cementitious materials, and colloidal transport of various radionuclides. The total exposure time (in some cases) is anticipated to be as long as 10 years. Soils and cementitious materials will be placed in the lysimeters and the releases will be determined from the leachates and extracted solid samples after environmental exposure. Liquid leachate will be continually gathered and analyzed.

Expected Benefit: This task is expected to provide actual site-specific Kd values in soil and cementitious materials and colloidal transport measurements for various radionuclides. It will provide additional information about long-term geochemical and transport phenomena that will be used to support the waste release and transport models used in the SDF, FTF, and HTF PAs.

FY2012: Completed the installation of the lysimeter and initiation of the sample collection program. [SRNL-STI-2012-00603]

FY2013: The concentrations of radionuclides and stable iodine (leached from either filter “pita pockets” or cementitious monoliths) were measured in the effluents from the field lysimeters. Key findings include:

• Cementitious saltstone monoliths (with and without reducing BFS) were spiked with Tc-99 and iodine, and loaded into the soil lysimeters. The concentration of iodine in the lysimeter effluent was very low or below detection limits. With respect to Tc-99 concentrations ranging from 2.7E-04 pCi/L to 9.5E-06 pCi/L were measured in the lysimeter effluents. These values represent between 10% and 50% of the total Tc-99 contained within the cementitious monoliths source leaching out and transporting through the soil column. The lysimeters containing saltstone with no BFS exhibited the largest fraction released. Because of uncertainties related to the technetium and iodine sample conditions at the beginning of leachate collection, these lysimeters have been capped.


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Figure 2.3-7: Ongoing Studies


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• A gamma suite of radionuclides (Cs-137, Co-60, Ba-133, and Eu-152) was spiked into cementitious saltstone (with and without reducing BFS), and loaded into the soil lysimeters. Cobalt-60 was the only gamma emitting radionuclide measured in the effluent. The highest concentrations were observed eluting from the lysimeters containing saltstone with no added BFS.

The concentrations of neptunium (Np) and plutonium (loaded into the lysimeters inside filter “pita pockets”) measured in the lysimeter effluents were very low or below the detection limit. [SRR-CWDA-2013-00121]

FY2014: The concentrations of radionuclides (leached from either filter “pita pockets” or cementitious monoliths) were measured in the effluents from the field lysimeters. [SRRA021685] Key findings include:

• The concentrations of plutonium in the effluents were below the detection limit for all lysimeters containing plutonium sources. Because plutonium has a high soil Kd, appreciable leaching was not expected.

• Lysimeters 29 and 30 have had measureable effluent concentrations of Np-237 corresponding to 0.07% and 4.44% of the initial activity added to the source. This is consistent with the higher mobility of pentavalent Np(V), which is the source in Lysimeters 29 and 30. It is also noteworthy that no breakthrough occurred from Lysimeters 31 and 32 which contained neptunium in the less mobile tetravalent state (Np(IV)).

• Cobalt-60 was the only gamma-emitting radionuclide measured in the effluent. The highest concentrations were observed eluting from the lysimeters containing saltstone with no added BFS. Note that Co-60 is characterized as a surrogate for nickel and cadmium.

FY2015: Key findings with respect to plutonium and cobalt transport are essentially the same as FY2014. Similarly, lysimeters 29 and 30 have had measurable effluent concentrations of Np-237 corresponding to 1.56% and 13.1% of the initial activity added to the source. This is consistent with the higher mobility of pentavalent Np(V), which is the source lysimeters 29 and 30. No breakthrough was apparent for lysimeters 31 and 32 which contained neptunium in the less mobile tetravalent state (Np(IV)). [SRRA021685-000007]

FY2016: Key observations for FY2016 are as follows:

• The concentration of Pu in lysimeter effluents were on the order of E-15 to E-13 mol/L. This was below the detection limit for the standard ICPMS measurement and a low-level radioanalytical technique was used to perform these measurements.

• Lysimeters containing NpO2NO3 sources have measurable breakthrough corresponding to 2.5% and 18.6% of the initial activity added to the source. The variability in these numbers is hypothesized to be caused by heterogeneous flow of water through the lysimeter. During work in FY2016, Np was also observed in the effluent of lysimeter 32, which contains a relatively insoluble NpO2 source. The


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observation of Np in the effluent from this lysimeter implies that the NpO2 is becoming oxidized and releasing Np(V) which can transport through the lysimeter with a relatively low Kd.

• Similar to previous years, Co-60 was measured in the effluent of all lysimeters containing gamma emitting radionuclides. However, the concentrations were much lower than previous sampling events. The majority of the Co-60 was released within the first two years of the experiment and concentrations are now close to detection limits. All cement and saltstone lysimeter sources contained higher concentrations of Co-60 in the effluent relative to a control with the gamma suite of radionuclides added directly to a filter.

• There is a high degree of variability in the amount of water flowing through each lysimeter. It is hypothesized this is due to heterogeneous flow of water through the soil and variations in the localized climate (i.e., wind and rain patterns) above the four-inch diameter lysimeter opening.

FY2017: Two reports were issued in September 2017. The first report (SRRA021685-000008) documented concentrations measured in field lysimeter effluents during FY2017.

Key findings from this report include:

• Lysimeters containing NpO2NO3(s) sources (Np in the +5 oxidation state) continue to show measurable Np breakthrough. The variability in these numbers is hypothesized to be caused by heterogeneous flow of water through the lysimeter. Starting in FY2016 and continuing into FY2017, Np was observed in the effluent of lysimeter 32 which contains a relatively insoluble NpO2(s) source. The observation of Np in the effluent from this lysimeter implies that the NpO2(s) is becoming oxidized and releasing Np(V) which can transport through the lysimeter with a relatively low Kd.

• Similar to previous years, Co-60 was measured in the effluent. However, the concentrations were much lower than previous sampling events. The majority of the Co-60 was released within the first two years of the experiment and concentrations are now close to detection limits. All cement and saltstone lysimeter sources contained higher concentrations of Co-60 in the effluent relative to a control with the gamma suite of radioisotopes added directly to a filter. It is unclear what is causing this enhanced mobility of a small fraction of Co-60 in the cement and saltstone sources.

• There is a high degree of variability in the amount of water flowing through each lysimeter. It is hypothesized this is due to heterogeneous flow of water through the soil and variations in the localized climate (i.e., wind and rain patterns) above the four-inch diameter lysimeter opening.

The second report (SRRA021685-000009) documented the detailed solid phase analysis of a field lysimeter with an emplaced colloidal PuO2(s) source.

Key findings from this report include:

• Consistent with previous lysimeter studies, both downward and upward migration of Pu was observed. A lower degree of upward migration is proposed to be due to the


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shorter time for Pu diffusion in the current lysimeter (3.4 years) compared with previous lysimeter studies (11 years).

• The downward transport of Pu in the current PuO2 lysimeter was greater than that observed for previous PuCl3, Pu(NO3)4, and Pu(C2O4)2 bearing lysimeters. This enhanced transport is hypothesized to be due to the colloidal nature of PuO2(s) which could result in enhanced transport.

• Desorption experiments using Pu contaminated soils retrieved from this lysimeter indicated conditional desorption distribution coefficients of log K = 4.4 ± 0.3 L/kg. There was no apparent difference between unfiltered and ultrafiltered samples during the desorption experiments indicating that either 1) colloids are not present in these samples as hypothesized or 2) colloids sorb strongly to the soil and do not desorb.

Lysimeter effluent testing in conjunction with solid phase analysis of the lysimeter cores and source material provides researchers with a robust data set specific to the SRS that can provide less ambiguous assignment of transport mechanisms, including Kd values, and bolster confidence in PA modeling assumptions. For example, values taken from the lysimeter solid phase analysis study provide strong support for vadose zone plutonium Kds that are one to two orders of magnitude higher than what is currently used in PA modeling at SRS.

FY2018: In FY2018, liquid samples will continue to be collected quarterly from the field lysimeters and transported to Clemson University. For cost savings purposes, only samples taken every other quarter (i.e., biannually) will undergo analysis. Effluent samples collected but not analyzed will continue to be stored at Clemson University and can be analyzed at a later date if funding becomes available. Since breakthrough has been detected for two lysimeters with Np sources, monthly effluent sampling and analysis will be performed for lysimeter 30 (containing a NpO2NO3(s) source) and lysimeter 32 (containing a NpO2(s) source) to better capture the radionuclide’s release and transport behavior. Solid phase analysis of lysimeter 41 containing an emplaced Pu(V) source is scheduled for FY2018. This analysis will help to ascertain what impact plutonium’s initial valence state plays in its migration behavior.

FY2019 through FY2022: Collect liquid samples from the lysimeter and transport to Clemson University for applicable analysis. The possibility exists that for any given sample event, sufficient rainfall may not have occurred and that leachate samples will not have accumulated such that there will not be data for that sample event.

Deliverable: Annual Leachate and Solid Phase Analysis Reports, as applicable

Expected Completion Date: After FY2022


Estimated Cost: FY2018 $41K (plus $75K funded under Tank Farm), FY2019 through FY2022 $0/yr


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2.3.2.2 Studies Related to Cementitious Materials Degradation Due to Radiation Damage

Description: Saltstone is a cementitious waste form. As such, damage to cementitious materials from radiolytic mechanisms must be understood. A literature search will be conducted to gain a better understanding of the potential degradation of cementitious materials exposed to radiation.

Expected Benefit: This activity is expected to produce a baseline of knowledge concerning cementitious degradation due to radiolytic mechanisms over long periods to inform degradation assumptions. The data provided from this effort will also inform the HTF and FTF modeling.

Deliverable: Technical Report

Expected Completion Date: To Be Determined



2.3.2.3 Closure Cap Long-Term Performance Description: This task involved research and development regarding the long-term performance of the closure cap as performed by the University of Virginia. The work is intended to derive analysis results to replace the current engineered closure cap baseline for the SDF as documented in WSRC-STI-2008-00244. Particular emphasis will be directed towards the following scope:

• Establishing infiltration rates through the bottom of the engineered closure cap system over time that can be utilized as an upper boundary condition to vadose zone flow and transport modeling.

• Determining uncertainty ranges in the infiltration rates that can be utilized in alternate infiltration scenarios for PA modeling.

• Providing recommendations for any improvements to the methods currently employed to degrade the performance of the sand drainage layer above the SDUs.

• Comparison to alternate calculation methods/codes or to measured/literature values to provide model support for baseline analysis.

Expected Benefit: This effort will validate assumptions in the PAs concerning the rate of closure cap infiltration as well as the behavior of the drainage layer above each SDF disposal unit.

FY2018: The technical report to be issued in FY2018 will make extensive use of the knowledge on hydraulic properties that has developed over the last decade, including the most recent data from DOE-LM. An extensive review of hydraulic properties of engineered cover soils will be conducted using the data collected by EPA, NRC, and DOE as well other information in the literature. This information will be compiled into a set of recommendations


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to describe expected hydraulic properties of each layer in the cover profile as well as the uncertainty (or range) for each of the properties. This "catalog of hydraulic properties" will be used as input to a coupled model used to predict leakage from the base of the composite barrier.

Hydraulic properties of the lateral drainage layer are of particular interest. A common assumption is that coarse-grained drainage layers will "silt in" over time as fines from overlying layers migrate downward. This common assumption frequently is not realized in the field, especially at depth in the vadose zone where hydraulic gradients are very low and bioturbation is limited or non-existent. For example, exhumations conducted during the field studies for NUREG CR-7028 demonstrated that engineered cover profiles with fine-over-coarse layers used as capillary breaks have remained intact during service, with no measurable ingress of fines. Similarly, natural analogs in eastern Washington show sharp contrasts between adjacent coarse- and fine-grained layers in the vadose zone that have been stable in natural sedimentary deposits for more than 10,000 yr. These analogs formed the basis for the sharp contrast in hydraulic properties employed in the analysis of cover systems at Hanford employing the Hanford Barrier design. In this project, a detailed review of the literature will be conducted to ascertain natural analogs of similar fine- over coarse-grained layers in fluvial deposits of the southeastern US. These analogs will be used to provide the technical justification for hydraulic properties used in the analysis for the engineered closure cap.




Estimated Cost: FY2018 $65K, FY2019 through FY2022 $0/yr

2.3.2.4 Hydraulic Degradation Mechanisms in SDU Concrete Description: This task is intended to establish key degradation mechanisms for SDU concrete based on long-term, realistic service conditions and incorporating credible, chronic exposure scenarios. For example, the majority of the interior SDU walls will be in direct contact with cured saltstone as opposed to the caustic saltstone pore solution, and therefore chronic exposure to pore solution chemistry would not be deemed a realistic service condition.

Expected Benefit: The current SDU concrete hydraulic degradation profile is represented as a linear degradation that occurs over a long period. The intent of the work is to gain an enhanced understanding of mechanisms potentially associated with the degradation of SDU concrete and utilize that information to recommend an alternative degradation profile to the currently utilized, and conservative, linear profile and is being conducted by Vanderbilt University. For example, while chemical species may diffuse into the SDU concrete, it is not known if said chemical diffusion ultimately results in reduced physical integrity.

FY2018: A final report is expected in FY2018 detailing the results of this study on concrete degradation rates and will provide recommendations that the effective saturated hydraulic conductivity of the degraded SDU vault concrete should be based on the relative thickness‐weighted geometric mean of the intact concrete and that degraded layers hydraulic conductivities be applied where the degraded layer is represented as backfill material.


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This report should lead to a more realistic and defensible estimate of the effective hydraulic conductivity in degraded vault concrete while remaining conservative. The impact of using the more realistic estimate of the effective saturated hydraulic conductivity for the degraded concrete should decrease the effective saturated hydraulic conductivity by up to several orders of magnitude in 1,000 years.





2.3.3 Emplaced Saltstone Sampling and Characterization Maintenance activities presented in this section will establish the research and development programs in order to measure properties of emplaced saltstone samples. In the past, saltstone samples collected from SDU 4 underwent testing. That testing indicated that alternative sample collection methodologies should be employed that may have less impact on the sample itself.

A formed core-sampling device similar to that described in Design and Testing of the Formed-Core Sampling System for Saltstone Facility Vaults has been constructed and tested at full scale. [SRNL-STI-2010-00167] The test included a comparison of hydraulic conductivity, porosity, and bulk density between the formed core simulant and cores drilled from a test apparatus. The results showed that there were differences in the hydraulic conductivity between the different sample types and this was attributed to the handling and storage of the samples. [SRNL-STI-2012-00551] The bulk density and the porosity of the two samples were comparable.

During the design and mock-up testing of the formed core sampling device, it was determined that this sampling technique for emplaced saltstone sampling was not practical and that a different technique would be required to collect an emplaced saltstone sample from a SDU. Based on this determination, an alternative sampling strategy has been developed. Full details of the sampling strategy are provided in the Saltstone Sampling and Analyses Plan, and key elements of the strategy are depicted in Figure 2.3-4. [SRR-SPT-2012-00049] The intent is to demonstrate a correlation in key properties, such as SHC, between saltstone prepared in the laboratory (using simulant waste) and saltstone emplaced into an SDU and subsequently retrieved via a core drilling process. In order to facilitate this correlation, the preparation and analyses of a number of intermediate sample sets were proposed to elucidate potential anomalies between the properties of the laboratory processed saltstone (using simulant waste) with saltstone prepared in the SPF (using actual waste from Tank 50). For example, retrieval of fresh saltstone samples from the SPF that were subsequently cured in a laboratory environment might highlight property variations resulting from actual and artificial curing environments. Tasks completed with respect to emplaced sampling are detailed below.


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2.3.3.1 Measure Physical Properties of Laboratory Prepared Saltstone Simulant Samples, Actual Tank 50 Salt Solution Samples, Saltstone In-Line Process Sample, and SDU Cell 2A Emplaced Core Sample

Description: Saltstone Engineering and the SRR Grout Subject Matter Expert developed a testing program that compared laboratory prepared saltstone samples with actual emplaced saltstone samples taken from SDU Cell 2A. This program was a multi-year effort involving the processing of saltstone samples in the laboratory, fresh saltstone collection from the SPF, optimization of the core drilling technique for emplaced grout, and eventual retrieval of emplaced grout and comparison with laboratory prepared saltstone with respect to key properties affecting the modeling. Additional information on the samples collected can be found in the Saltstone Sampling and Analyses Plan, SRR-SPT-2012-00049.

Expected Benefit: This testing program was intended to provide confidence that the laboratory prepared simulated saltstone grout has similar properties to the emplaced saltstone grout and that modeling assumptions can be made on laboratory-prepared samples.

FY2013: Details regarding activities accomplished in FY2013 towards eventual retrieval and characterization of emplaced saltstone grout were detailed in the Update of Fiscal Year 2013 Activities Related to SDU Sampling and Analyses. [SRR-SPT-2013-00044] Key achievements were as follows, and are additionally highlighted in Figure 2.3-8.

• Installation of the inline sampler downstream of the SPF grout hopper. A total of 36 sample molds were filled from the sampler and subsequently transferred to SRNL for curing. Half of the samples were being cured at room temperature and half according to a simulated SDU temperature/humidity environment.

• Retrieval of two 3-liter samples from Tank 50 which will be used as needed for elucidating potential anomalies associated between the properties of laboratory prepared and emplaced saltstone. [SRR-SPT-2012-00049] Tank 50 solution will be utilized (at a later date if needed) for processing saltstone samples in the laboratory. A determination of need will depend on the comparison of properties for samples retrieved from the SDU and samples prepared in the laboratory using simulant (non-radioactive) Tank 50 salt solution.

• Preparation of 36 saltstone samples in the laboratory utilizing simulated (non-radioactive) Tank 50 salt solution. Half of the samples were cured at room temperature and half according to a simulated SDU temperature/humidity environment. Properties of the laboratory prepared samples will be compared to those of the actual saltstone samples to be retrieved from SDU Cell 2A.

• Filling of a B-25 container with saltstone simulant for core drilling mock-up. Approximately 700 gallons of grout were processed using the scaled grout mixer at SRNL, and poured into a B-25 container in daily lift heights of approximately eight to nine inches to mimic actual SDU emplacement. Saltstone from the scaled mixer was also poured into cylindrical molds to enable property comparison between cast and cored samples. The grout was cured for 60 days prior to core drilling.


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• Mock-up of SDU coring process (using grout poured into the B-25 container) to optimize the coring and retrieval process. The mock-up was set up to simulate the coring of an SDU, and incorporated five feet of vapor space and approximately one foot of grout above the desired sample location. Both wet and dry coring were assessed and wet coring deemed preferable based on the ease of extraction of the cores from the coring bits. The wet drilled cores contained multiple through-thickness fractures predominantly associated with lift height joints. The longest intact wet cored samples retrieved were approximately eight to nine inches. Retrieval of the cores utilizing a specially designed core extraction tool proved unsuccessful.

• Saturated hydraulic conductivity and compressive strength were measured on cast and core drilled samples from the mock-up of SDU coring process. The compressive strengths of the cored samples (500 psi) were approximately 50% of those measured for the cast samples. No damage from coring that would account for the lower strengths was observed on the samples prior to testing. In contrast, the saturated hydraulic conductivities (a key input for contaminant transport models) of cored and cast samples all measured at approximately 2.0E-09 cm/s.


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Figure 2.3-8: Emplaced Saltstone Testing Strategy (Detailed in SRR-SPT-2012-00049)


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FY2014: Samples from FY2013 that were either retrieved from the SPF, or prepared in the laboratory, have been maintained in their respective curing environments. These samples will continue to cure until actual cores have been retrieved from the SDU after which each saltstone sample type will be analyzed with respect to key properties, such as SHC, density, porosity, and technetium oxidation state and leaching behavior.

Cores retrieved from the FY2013 mock-up (Phase 1) were analyzed in FY2014 with respect to SHC and compressive strength, and compared to samples cast into molds using the same saltstone materials. Compressive strength, although not an input to the SDF PA models, was measured since it is a simple, economical, and well-defined testing method that would be expected to emphasize any mechanical damage resulting from core-drilling. Tc-99 leaching behavior was not measured directly but potential impacts to this measurement were inferred from visual examination. In summary, the SHC values for both core-drilled and cast samples were approximately equal and measured (on average) 1.8E-09 and 2.1E-09 cm/sec, respectively. However, compression testing revealed that the strength of the core-drilled samples was (on average) approximately half the strength of the cast samples (550 psi compared to 1,000 psi). Ultimately, visually observed surface scoring (from the rotating drill bit) were considered likely physical features that could yield lower compressive strengths for the core-drilled samples.

A continuation of the simulant saltstone core-drilling mock-up (Phase 2) was also conducted in FY2014 with key objectives defined as demonstrating a viable core extraction process, optimizing core-drilling techniques to reduce sample damage (including through-thickness fractures and surface scoring), and determination of water ingress into the core interior during the wet drilling process. [SRR-CWDA-2014-00059]

A simple retrieval tool, essentially consisting of a plastic tube and a retainer basket, was developed in-house, and proved highly successful for retrieving full-length (3.5-foot) core samples (divided into multiple smaller samples due to through-thickness fractures). With respect to optimization of the core-drilling process, particular emphasis was placed on reducing drill wobble that was observed in the FY2013 mock-up and determining the optimum water flow for removing cutting fines from the cutting surface of the diamond core bit, and for preventing the binding of core samples in the core bit and core extension barrels. Drill wobble was tentatively presumed to be a factor in through-thickness fracturing of the cores, and circumferential scoring of the core surface, observed during the FY2013 mock-up tests. While improving the stability of the drill and utilizing new extension barrels was visually determined to significantly reduce drill wobble, both scoring and through-thickness fractures of the samples remained. To investigate potential water ingress to the core interior during wet drilling, moisture contents were measured for samples that were wet and dry core-drilled. For both sample types the moisture content was 36 to 37%, indicating that wet core-drilling does not cause water ingress and will not impact the measurement of contaminant leachability.

FY2015: While the FY2014 mock-up activities proved highly successful and served to resolve most of the issues highlighted in the FY2013 mock-up, further testing (Phase 3) was required in part due to an increase in the fill height limit of the SDUs. When the mock-up plan was conceived the maximum SDU fill height was 18.5 feet. Samples would have to be retrieved from the 16-foot elevation in SDU Cell 2A, and thus mock-up testing assumed an overlying


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grout layer of approximately 3 feet. However, the SDU fill heights were increased to 21.5 feet, and thus the overlying grout layer has increased to approximately 6 feet. The simulant saltstone monolith utilized in both FY2013 and FY2014 mock-up activities has a maximum grout depth of 3.5 feet, and thus additional testing was required in which the drilling and retrieval processes could be demonstrated to a depth of 6 feet or greater. While it was not anticipated that wet core-drilling to a depth of at least 6 feet would be problematic, the subsequent extraction, handling, packaging, and transport of samples potentially 6 feet in length required further investigation. The final mock-up phase was conducted at the beginning of calendar year (CY) 2015, and demonstrated core-drilling and sample retrieval to a depth of 6 feet below the grout surface. [SRR-CWDA-2015-00002] Utilizing the drilling and extraction techniques optimized in the three mock-up phases, core samples were successfully extracted from SDU Cell 2A in April and May of 2015. SRR extracted approximately 191 inches of 2-inch diameter core material from the SDU. The core material was inerted after extraction and transported to SRNL for storage (in an inert atmosphere) and analyses. [SRR-CWDA-2015-00066] FY2016: As indicated in the previous text, multiple cores were extracted via a wet core drilling process in April-May 2015 approximately 20 months after the saltstone of interest was processed in the SPF and subsequently emplaced in SDU Cell 2A. The retrieved samples were stored in an inert N2 environment prior to and during the property testing which commenced in October 2015 and was completed in March 2016. Characterization was predominantly conducted on laboratory-prepared, non-radioactive, simulant saltstone samples and the radioactive samples extracted from SDU Cell 2A. The physical property data for SDU-emplaced and laboratory-prepared samples is summarized in the SDU Cell 2A Core Sampling Report (SRR-CWDA-2016-00051), with values for SDF model inputs provided in the report where applicable.

Physical properties such as bulk density, porosity, and water content highlight differences between the field and laboratory samples resulting from the variability of bulk processing and emplacement in comparison to the consistent preparation of laboratory samples in a controlled environment. Indeed, the SDU Cell 2A cores indicated lower bulk densities and higher porosity (both total and permeable) in comparison to the laboratory-prepared, simulant samples. It was postulated that SPF flush water additions during and after processing in the SPF were likely contributors to the lower density and higher porosity values. It was therefore surprising that the water loss (upon heating to 110 °C) for the SDU Cell 2A cores and the laboratory-prepared samples were almost identical. This result is tentatively, and partially, attributed to the handling of samples in an inert N2 chamber with essentially zero humidity.

While physical differences were apparent between the sample types, the key concern is whether those differences impact properties relevant to the long-term performance of saltstone, in particular its ability to retard the spread of radioactive contaminants to the surrounding environment. In view of the higher permeable pore volume in the SDU Cell 2A cores, there was a concern that this would manifest as higher permeability (or SHC) for the SDU Cell 2A cores in comparison to the laboratory-prepared samples. However, despite the fact that two of the six SDU Cell 2A cores analyzed indicated physical damage associated with the core drilling process, the mean SHC was shown to be less than 1.6E-09 cm/sec (with a highest absolute value of 4.4E-09 cm/sec). For demolded, defect-free, laboratory samples the SHC was


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determined to be less than 1.0E-09 cm/sec, which corresponds to the lower detection limit of the permeameter used for the testing. The SDF transport modeling utilizes a value of 6.4E-09 cm/sec, and it is particularly noteworthy that all six of the tested SDU Cell 2A cores indicated SHCs less than this value despite the aforementioned damage imparted on some samples by core drilling, and in addition to the fact that the samples were extracted from three separate SDU locations and from heights between 15 and 17 feet above the SDU floor. This latter detail attests to an inherently low SHC for saltstone irrespective of processing variability and location within an SDU.

The final properties of interest with respect to the SDU Cell 2A samples were related to their ability to restrict the migration of radionuclides from the saltstone monolith to the surrounding environment. For this study Tc-99, Sr-90, and I-129 were evaluated noting that testing was limited to the actual SDU Cell 2A cores; the principal objective was to confirm the radionuclide transport properties assumed for SDF transport modeling. Testing involved the addition of 1 gram of ground SDU Cell 2A core to 10 mL of leachate, with (oxic) and without (anoxic) dissolved oxygen (DO), followed by seven day mixing to facilitate “pseudo” equilibrium of the solid-liquid phases. With respect to Tc-99, under reducing (or anoxic) conditions the SDF transport model assumes that Tc-99 leaching is controlled by the solubility of reduced TcO2.xH2O phases. Tc-99 concentrations in the DO-depleted (anoxic) leachate after seven days confirmed this assumption. The pH and Eh of the leachate predicted TcO2.xH2O as the most thermodynamically stable species, and the mean Tc-99 concentration (for six SDU Cell 2A cores) of 2.2E-08 mol/L (at pH 10.5) was synonymous with the solubility limits of hydrated Tc-oxide phases. While the SDF transport model utilizes a fixed solubility of 1.0E-08 mol/L for Tc-99 (slightly lower than that measured for the SDU Cell 2A cores), sensitivity analysis at an order of magnitude higher Tc-99 concentration (i.e., 1.0E-07 mol/L) indicated no significant impact with respect to the timing or magnitude of the peak dose exposures in the surrounding environment. Surprisingly, for tests in which the leachate contained DO (oxic) the Tc-99 solubility was relatively unchanged (2.3E-08 mol/L) which is perhaps indicative of the stability of the reduced TcO2.xH2O phases and/or the slow redox transformation kinetics as the aqueous system transitions from reducing to oxidizing.

Both Sr-90 and I-129 are considered highly mobile contaminants and their retention is assumed to be predominantly controlled by adsorption to the interior surfaces of saltstone, such as capillary pores (i.e., Kd-controlled). Adsorption properties in aqueous systems are characterized by measuring how the contaminant of interest partitions (distributes) between the solid and liquid phases of the system. The propensity for adsorption is specified as a distribution (or partitioning) coefficient (Kd) (mL/g) which is the ratio of the contaminant concentration (or activity for radionuclides) associated with the solid (pCi/g) to the contaminant concentration associated with the liquid (pCi/mL). A low Kd value (i.e., close to 0 mL/g) indicates that a contaminant partitions readily from the solid into the solution and is therefore highly mobile; whereas a high Kd value indicates a less mobile contaminant. The Kd assumes that the aqueous system has reached equilibrium but given the slow kinetics associated with many chemical transitions, and therefore the excessively long timeframes that would be required to achieve equilibrium in laboratory experiments, Kd may be referred to as a distribution ratio (Rd) which acknowledges that full equilibrium may not have been attained in shorter duration tests (e.g., the seven-day test for this study).


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With respect to Sr-90, the Kd (or Rd) in both oxic and anoxic environments was greater than that assumed in the SDF transport model (i.e., 15 mL/g) and as such demonstrate that the assumptions used in the SDF model are conservative. In contrast, the calculated Kd for I-129 in oxic and anoxic environments was <1 mL/g and could potentially be designated as zero inferring that all of the I-129 contained in the SDU Cell 2A cores was partitioned to the leachate. Indeed, for some cores, the Kd was calculated to be a negative value which is not feasible since it implies that I-129 is being generated during testing. Negative Kd values are artifacts of a reverse Kd (or desorption) technique which required analytical measurement of the total I-129 contained in an untested SDU Cell 2A and the I-129 partitioned to the liquid phase (after the seven-day test) both of which are subject to analytical uncertainties due to the inherently low concentration of I-129 contained in saltstone. A zero Kd also seems extremely unlikely since at least a small proportion of the I-129 is expected to reside within the impermeable pore volume of saltstone. In addition, the sensitivity models show that even when assuming a Kd value of 0 mL/g in reducing saltstone (i.e., complete mobilization of iodine), the release of I-129 does not increase by more than a factor of two, such that the peak dose contribution from I-129 only increased from about 11 mrem/yr to about 17 mrem/yr within 10,000 years of facility closure. This is because with relatively low iodine Kd values in saltstone, the release of I-129 is more a function of flow rates, rather than a function of the Kd. Since the initial SHC is expected to be lower than currently modeled (based on the data in this report), the lower SHC is anticipated to decrease flow rates through saltstone, and the actual change to the modeled peak dose for I-129 should be less significant.

With respect to the objectives of the Saltstone Sampling and Analyses Plan (SAP) (SRR-SPT-2012-00049) and the data derived from this work has demonstrated the following:

1. While minor physical differences (i.e., density and porosity) may exist between field- and laboratory-processed samples, the measured differences were not impactful with respect to key contaminant transport properties, in particular SHC. As such, there can be greater confidence that studies of laboratory samples can provide sufficient insight into the inherent physical properties of field processed and emplaced saltstone.

2. The assumptions currently made in the SDF transport model with respect to SHC, and radionuclide leaching behavior, have predominantly been affirmed. In particular, the data from this study supports that Tc-99 is reduced in saltstone, and that modeling Tc-99 leaching behavior as solubility limited (in a reduced, high pH environment) is appropriate. With respect to the adsorption (or rather desorption) characteristics of I-129, while the reverse Kd was less than 1 mL/g, sensitivity analysis indicated that even a zero Kd (0 mL/g) would not significantly impact the magnitude or timing of the peak dose in the surrounding environment.

In addition to the aforementioned analysis, the leaching characteristics of Tc-99 were also evaluated in FY2016. Using EPA Method 1315, Tc diffusivities was determined for three intact SDU Cell 2A cores; the values ranged from 5.2 to 6.4E-11 cm2/sec. As with the aforementioned simulant samples, the effective diffusivities of poorly adsorbed nitrates were approximately two orders of magnitude higher than the Tc-99 diffusivities. Measurement of I-129 concentrations in the EPA method leachates and subsequent calculation of the effective


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diffusivities of radioiodine from the SDU Cell 2A is currently pending. This data was presented in SREL Doc. R-17-0004.

Two SDU Cell 2A cores were also evaluated using the recently developed DLM test. With respect to SHC, the Tc-spiked sample (refer to Section 2.3.1.2) has indicated a relatively consistent value that fluctuates ranging between 5E-10 and 2E-09 cm/sec. In contrast, one of the SDU Cell 2A cores indicated a high initial SHC of 1E-08 cm/sec which subsequently dropped <5E-10 cm/sec after 1.5 pore volume exchanges. The second SDU Cell 2A sample started with an SHC of 2E-09 cm/sec which dropped to <1E-10 cm/sec with less than 1 pore volume exchanged. Initial Tc-99 concentrations for the SDU Cell 2A cores are approximately 2E-07 mol/L.

FY2017: Using EPA Method 1315 the diffusivities of I-129 and Cs-137 diffusivities for the three SDU Cell 2A cores ranged 1.0E-08 to 5.5E-09 cm2/sec and 1.1 to 4.9E-10 cm2/sec, respectively. The leaching behavior for I-129 was similar to that observed for poorly adsorbed NO3

- as indicated by the high effective diffusivities for both components [SREL Doc. R-17-0005].

Two intact SDU Cell 2A cores have been subjected to DLM testing. Table 2.3-1 provides pertinent data for each of the SDU Cell 2A samples.

Table 2.3-1 DLM-Derived Data for SDU Cell 2A Samples.

SDU Sample PV

SHC (cm/sec) Tc-99 (M) Cs-137 (M) NO3- (M)

Start Current Start Current Start Current Start Current

1 1.5 2.3E-09 9.4E-11 2.1E-07 9.5E-08 7.4E-09 2.3E-09 9.5E-01 3.0E-01

2 8.0 1.2E-08 5.3E-10 2.5E-07 4.1E-08 6.0E-09 2.5E-10 4.1E-01 2.1E-02

The sample exhibiting the lowest SHC of approximately 1E-10 cm/sec was halted after 1.5 pore volumes had been exchanged. The second SDU Cell 2A sample continues to be evaluated and approximately 8 pore volumes have been exchanged for the sample. At the time of test cessation, the SDU Cell 2A with low flow rate indicated cumulative Tc-99, Cs-137, and NO3

- percentages leached of 4.8, 6.7, and 34.5 %, respectively. The final SHC was measured at 9.4E-11 cm/sec. It is noteworthy that the SDF PA utilizes a saltstone SHC of 6.4E-09 cm/sec which is likely conservative based on the data measured for the SDU Cell 2A cores. After the exchange of 8 pore volumes, the second SDU Cell 2A sample indicates cumulative Tc-99, Cs-137, and NO3

- percentages leached of 11.7, 11.7, and 77.4 %, respectively [SREL Doc. R-17-0005]. Figure 2.3-9 indicates SHC data as a function of pore volumes exchanged for the SDU Cell 2A samples, and demonstrates a general decrease in SHC as the number of pore volumes exchanged increases. Figures 2.3-10 and 2.3-11 provide the leachate concentration (Tc-99, Cs-137, and nitrate) for both SDU Cell 2A samples. Figure 2.3-12 indicates the pH and Tc-99 concentrations for the SDU Cell 2A cores; the graph also provides the thermodynamically-predicted relationship between Tc compound solubility and pH. The Tc-pH relationship indicates that the Tc in the cores is likely in the form of a reduced TcO2.2H2O phase.


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Figure 2.3-9: SHC Versus Pore Volumes Exchanged for SDU Cell 2A Cores

Figure 2.3-10: Leachate Concentration Data for SDU Cell 2A (SDU2A-0931-C-1-U-5)

1E-12

1E-11

1E-10

1E-09

1E-08

1E-07

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0

SDU2A-0931-C-1-U-2

SDU2A-0931-C-1-U-5


SHC

(cm

/sec

) 5.3E-10 cm/sec

9.4E-11 cm/sec

1E-10

1E-09

1E-08

1E-07

1E-06

1E-05

1E-04

1E-03

1E-02

1E-01

1E+00

1E+01

1E+02

0.0 0.5 1.0 1.5 2.0

Leac

hate

Con

c. (

mol

/L)

NO3- (0.29 mol/L)

137Cs (2.3E-09 mol/L)

99Tc (9.5E-08 mol/L)

SDU Cell 2A Core - SDU2A-0931-C-1-U-5Approx. pour date: Dec. 2013



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Figure 2.3-11: Leachate Concentration Data for SDU Cell 2A (SDU2A-0931-C-1-U-2)

Figure 2.3-12: Leachate pH and Tc Concentrations for SDU Cell 2A DLM Samples

1E-10

1E-09

1E-08

1E-07

1E-06

1E-05

1E-04

1E-03

1E-02

1E-01

1E+00

1E+01

1E+02

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0∑Pore Volumes Exchanged

Leac

hate

Con

c. (

mol

/L)

NO3- (0.02 mol/L)

137Cs (2.5E-10 mol/L)

99Tc (4.1E-08 mol/L)

SDU Cell 2A Core - SDU2A-0931-C-1-U-2Approx. pour date: May 2014

1E-09

1E-08

1E-07

1E-06

1E-05

1E-04

1E-03

1E-02

1E-01

1E+00

1E+01

9 10 11 12pH

Solu

bilit

y (m

ol/L

)

TcO2.1.6H2O

TcO2.2H2O

KTcO4

NaTcO4.4H2O


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On March 23, 2017, the NRC issued Technical Review Report Saltstone Waste Form Hydraulic Performance (ML17018A137), which was related to Monitoring Area (MA) 3 (Waste Form Hydraulic Performance) and MA 10 (Performance Assessment Model Revisions) from that Monitoring Plan. The NRC Technical Review Report (TRR) was based on the NRC staff review of 14 primary DOE documents. In the NRC TRR, the technical staff recommended to close MFs in MA 3, but, did not recommend closing any MFs in MA 10.

In that TRR, the NRC technical staff recommended closing MF 3.01 (Hydraulic Conductivity of Field-Emplaced Saltstone) under both Title 10, Code of Federal Regulations (10 CFR) Part 61 Performance Objective (PO) §61.41 (Protection of the General Population from Releases of Radiation) and PO §61.42 (Protection of Individuals from Inadvertent Intrusion) (i.e., two specific MFs). The reason for that recommendation was that the NRC determined that research results provided by the DOE were adequate to support the assumed initial hydraulic conductivity of field-emplaced saltstone in both the FY2013 SDF SA and FY2014 SDF SA. [SRR-CWDA-2013-00062, SRR-CWDA-2014-00006] The NRC technical staff recommended closing MF 3.02 (Variability of Field-Emplaced Saltstone) under both 10 CFR Part 61 PO §61.41 and PO §61.42 (i.e., two specific MFs). The reasons for that recommendation were: (1) the DOE final research results provided to the NRC in early Calendar Year 2017, in particular, the measured properties of saltstone core samples, provided significant insight into variability of field-emplaced saltstone; (2) the NRC determined that the production, placement, and curing conditions that could cause significant variability in saltstone performance were well-controlled by the DOE and the NRC did not expect the variability in those conditions to result in significant variability of field-emplaced saltstone; and (3) the NRC determined that the DOE process to evaluate variability due to potential future changes was adequate to assess and control saltstone variability.

The NRC technical staff recommended closing MF 3.04 (Effect of Curing Temperature on Saltstone Hydraulic Properties) under both 10 CFR Part 61 PO §61.41 and PO §61.42 (i.e., two specific MFs). The NRC determined that the DOE research results since the DOE 2009 SDF PA demonstrated that the effects of curing conditions were adequately accounted for in the assumed initial hydraulic conductivity and effective diffusivity values.

Based on the TRR recommendations, the NRC determined that the DOE has provided sufficient information to close those six specific MFs in the Monitoring Plan. In FY2017, the NRC issued a memo, Closure of Monitoring Factors in the 2013 U.S. Nuclear Regulatory Commission Saltstone Disposal Facility Monitoring Plan (Docket No. PROJ0734) (ML17097A351) to document closing these six MFs associated with saltstone.

FY2018: Continue DLM analysis of SDU Cell 2A core(s).

Deliverable: Core Sampling Technical Report (SRNL-L3100-2015-00073) – Complete Core Sampling Technical Report (SRNL-STI-2106-00106) – Complete Core Sampling Evaluation (SRR-CWDA-2016-00051) – Complete FY2017 Report (SREL Doc. R-17-0005) – Complete

Completion Date: FY2022



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2.3.4 To Be Determined Out-Year Testing Description: For FY2018 and beyond, testing has not been finalized. However, for budget purposes, estimated costs are included.

Responsibility: SRR WDA Estimated Cost: FY2019 $655K, FY2020 $530K, FY2021 through FY2022 TBD


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3.0 F-AREA AND H-AREA TANK FARMS

3.1 Tank Farm Performance Assessment Annual Maintenance Activities DOE M 435.1-1 requires the ongoing maintenance of all PAs. This maintenance involves a series of activities that must be performed on an ongoing or annual basis. The activities in this section represent those activities that will be required annually in support of the FTF and HTF PAs regardless of the status of any ongoing or future PA revisions.

3.1.1 Maintain Tank Farm Performance Assessment Control through Unreviewed Waste Management Question Process

Description: Similar to the process set up for evaluating disposal related questions in SDF, a UWMQ process was established for waste tank closure activities. The UWMQ process consists of providing UWMQEs of proposed activities or new information to ensure that the assumptions, results, and conclusions of the approved PA, CA, and SAs remain valid.

If identified through the UWMQ process that a proposed activity or new information is outside the bounds of the approved NDAA Section 3116 Basis Document, PA, CA, or SAs, new SAs are prepared to update the technical baseline. UWMQEs and SAs will continue to be required throughout the life of the facility. For planning purposes, the estimated cost assumes that four UWMQEs will be prepared each year in the out-years. The estimated cost does not reflect the cost of any emergent Tank Farm SAs. Currently planned SAs are captured in Section 3.2.4.

Deliverable: Provide UWMQEs and UWMQ procedure support, as needed to support closure of FTF and HTF




3.1.2 Prepare Annual Performance Assessment Maintenance Program Implementation Plan

Description: The purpose of the LW PA Maintenance Program is to confirm the continued adequacy of the current PA and SAs, and to increase confidence in the results. Every year the annual LW PA Maintenance Program FY Implementation Plan is prepared and provided to DOE-SR. Plan preparation will include review of outstanding PA and SA comments and recommendations (noted in Sections 3.2.2 and 3.2.3). The Implementation Plan will outline planned work for each FY. The cost of preparing the Implementation Plan will be shared between SDF and the Tank Farms. See Section 2.1.2 for SDF maintenance activities.

Deliverable: Issue a FY LW PA Maintenance Program Implementation Plan

Expected Completion Date: 1Q-2QFY (issued annually)




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3.1.3 Provide General Technical Support on Tank Farm Performance Assessment Issues

Description: This task is to provide general technical and programmatic support on Tank Farm PA and SA issues, NRC activities, and other regulatory issues that affect waste tank closure. Activities include testing and research activity support, general project support, review of annual groundwater monitoring data, supporting NRC on-site observation visits and technical reviews, and development of resolution path forward for NRC open items. Tier 2 Closure Plans are developed for each waste tank prior to closure activities and include reviews of actual tank residual impacts on long-term conditions. Research activity support includes monitoring of research done by outside agencies (e.g., academic research, Hanford activities) as well as research performed on-site (e.g., SRNL, SREL). These activities also include support on interactions with SCDHEC, SRS CAB, LFRG, National Academy of Sciences, and other regulatory and stakeholder bodies.

Deliverable: Provide ongoing technical support on regulatory and policy issues/activities affecting waste tank closure activities




3.1.4 Develop and Maintain Performance Assessment Model Archive and Revision Control

Description: This task established software and hardware resources for archiving development and final PA modeling files to a read-only storage medium. In FY2014, capital infrastructure improvements were enacted on the site network, allowing for faster communication between SRNL’s high performance computing network and SRR WDA servers. This improvement increased the rate for file transfers between the two systems. FTF and HTF modeling files (for both PORFLOW and GoldSim) were copied to electronic storage devices. The storage devices are maintained onsite by SRR WDA, within a cipher-locked facility. The properties of the electronic files were set to read-only. Copies of files can be provided upon request. As needed, additional storage devices will be purchased to provide sufficient disk space for maintaining a record of all related model files.

Deliverable: Establish process (completed in FY2014) and maintain after implementation




3.2 Tank Farm Performance Assessment Development/Revisions The FTF and HTF PAs provide the technical basis and results to be used in subsequent documents to demonstrate compliance with performance objectives of Licensing Requirements for Land Disposal of Radioactive Waste, Radioactive Waste Management Manual as required by NDAA Section 3116, Federal Facility Agreement for the Savannah River Site (FFA), Standards for


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Wastewater Facility Construction, and Proper Closeout of Wastewater Treatment Facilities. [10 CFR 61, DOE M 435.1-1, WSRC-OS-94-42, SCDHEC R.61-67, SCDHEC R.61-82]

3.2.1 Prepare Out-Year F-Area Tank Farm Performance Assessment Revisions Description: In March 2012, following issuance of the Section 3116 Determination for FTF Closure (DOE-WD-2012-001), DOE approved the Tier 1 Closure Plan for FTF (SRR-CWDA-2010-00147) including its referenced FTF PA, Revision 1 (Tier 1 authorization letter, DOE_03-28-2012, was received March 28, 2012). The FTF PA, Revision 1, has been issued and implemented. A future revision of the FTF PA will be scheduled as required and agreed upon by DOE. The current FTF PA will be revised when warranted, but for estimating purposes, the next revision will be scheduled starting in FY2027 and completing in FY2028 (in support of Tank 4 operational closure, the next scheduled FTF tank closure). Unless otherwise noted in the FTF PA, the future FTF PA revision will include the following items at a minimum:

• Updated modeling to make the PA more consistent with the more current modeling approaches (i.e., at a minimum, consistent with the HTF PA);

• Analyses and results contained in all SAs that have been completed to date; • Analyses and results of all UWMQEs completed to date; • Consideration of new information generated through research and development; • Changes in site future land use plans or closure plans; and • Changes to PA guidance documents requirements.

Future FTF PA revisions will also consider the following:

• LFRG open items for the following four LFRG criteria: 3.1.6.5, 3.1.8.1, 3.1.8.2, and 3.1.8.3 (LFRG_08-13-2008);

• Comment Responses to SCDHEC and EPA on Revision 1 of the FTF PA (SRR-CWDA-2011-00164, SRR-CWDA-2011-00175);

• Responses to RAIs posed by the NRC (SRR-CWDA-2011-00054); • NRC recommendations in the U.S. Nuclear Regulatory Commission Plan for

Monitoring Disposal Actions Taken by the U.S. Department of Energy at the Savannah River Site F-Area and H-Area Tank Farm Facilities in Accordance with the National Defense Authorization Act for Fiscal Year 2005 (ML15238A761), as discussed in detail in Section 3.3; and

• Information generated to support other PAs and SAs. • Incorporation of applicable TRRs from the NRC.

Furthermore, the future FTF PA revisions shall be in alignment with the most current revision of the LW System Plan. A revision of the FTF PA should be approved prior to the next tank closure action in the FTF.




Estimated Cost: FY2018 though FY2022 $0/yr


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3.2.2 Prepare Out-Year H-Area Tank Farm Performance Assessment Revisions Description: The HTF PA was submitted for DOE review via an LFRG review team in November 2010. Revision 1 of the HTF PA, incorporating FTF PA lessons learned and comments on HTF PA Revision 0, was dated November 2012. DOE’s Draft NDAA Section 3116 Basis Document for HTF was prepared in FY2013 and was provided, along with the HTF PA Revision 1, to the NRC to initiate HTF NDAA Section 3116 Consultation in FY2013. Final HTF PA approval and implementation was achieved in FY2015.

A future revision of the HTF PA will be scheduled as required and agreed upon by DOE. The HTF PA will be revised when warranted, but for estimation purposes, the next revision will be scheduled starting in FY2020 and completing in FY2021 (in support of Tank 15 operational closure, the next scheduled HTF tank closure). Unless otherwise noted in the HTF PA, the future HTF PA revision will include the following items at a minimum:

• Analyses and results contained in all SAs that have been completed to date; • Analyses and results of all UWMQEs completed to date; • Consideration of new information generated through research and development; • Changes in site future land use plans or closure plans; • Changes to PA guidance documents requirements; and • Modeling improvements as identified in the Quality Assurance report for the HTF PA

(SRR-CWDA-2012-00070).

Future HTF PA revisions will also consider the following:

• Comment Responses to SCDHEC and EPA on Revision 0 of the HTF PA (SRR-CWDA-2011-00135, SRR-CWDA-2012-00166);

• NRC recommendations from the Technical Evaluation Report for H-Area Tank Farm Facility, Savannah River Site, South Carolina (ML14094A496) and addressed within the Nuclear Regulatory Commission’s H-Tank Farm Technical Evaluation Report’s Recommendations – Department of Energy’s Activity Summary Matrix (SRR-CWDA-2014-00080);

• NRC recommendations in the U.S. Nuclear Regulatory Commission Plan for Monitoring Disposal Actions Taken by the U.S. Department of Energy at the Savannah River Site F-Area and H-Area Tank Farm Facilities in Accordance with the National Defense Authorization Act for Fiscal Year 2005 (ML15238A761), as discussed in detail in Section 3.3; and

• Information generated to support other PAs and SAs. • Incorporation of applicable TRRs from the NRC.




Estimated Cost: FY2018 through FY2019 $0/yr, FY2020 through FY2021 $500K/yr, FY2022 $0K


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3.2.3 Tank Farm Special Analyses Description: SAs are performed to evaluate the significance of new information or new analytical methods to the results and associated conclusions of a PA. As waste tanks and ancillary equipment are cleaned, final residual inventories will be used to update the PA fate and transport modeling, allowing for evaluation of the difference between the projected and final waste tank inventories to determine if the results and conclusions of the PA and supporting SAs remain valid.

During FY2016, an HTF Type I and Type II Tank SA document was prepared and issued. [SRR-CWDA-2016-00078] This SA provides additional information that will be used to inform decisions regarding HTF Type I and Type II Tank closure documents. The FY2016 HTF SA included the following:

1) HTF Type I and Type II Tank radiological and chemical inventories were updated. 2) HTF Base Case model inputs were updated and documented. 3) PORFLOW modeling runs were carried out using the HTF Base Case with updated

inventory and model inputs. 4) New sensitivity analyses were performed on areas of interest (e.g., inventory

variability, grout hydraulic performance) using the HTF PORFLOW model and HTF GoldSim model.

Currently, there are no SAs currently planned to be prepared in the period covered by this LW PA Maintenance Program.

Deliverable: HTF Type I and Type II Tank SA (SRR-CWDA-2016-00078) – Complete

Completion Date: FY2016 (HTF Type I and Type II Tank SA) – Complete



3.3 Tank Farm Performance Assessment Testing & Research Activities This section of the LW PA Maintenance Program Implementation Plan contains PA-related testing and research activities identified as part of the ongoing maintenance of the FTF and HTF PAs. The PA testing and research discussion within this section is intended to address combined testing and research activities for both FTF and HTF. No testing and research activities unique to a specific Tank Farm have been identified at this time.

Issuance of the FTF PA and the Basis for Section 3116 Determination for Closure of F-Tank Farm at the Savannah River Site (DOE/SRS-WD-2012-001) occurred in FY2012. In the U.S. Nuclear Regulatory Commission Planned Monitoring Activities for F-Area Tank Farm at the Savannah River Site (ML12345A322), the NRC made recommendations, with respect to the various monitoring factors identified by the NRC, for DOE to consider during maintenance and monitoring of the FTF PA (documented in Appendix A of ML12345A322). Subsequent to issuance of the FTF Monitoring Plan, the NRC issued an HTF TER. The recommendations included in the HTF TER and associated transmittal letter (provided subsequent to the FTF Monitoring Plan) were considered by DOE within the Nuclear Regulatory Commission’s H-Tank Farm Technical


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Evaluation Report’s Recommendations – Department of Energy’s Activity Summary Matrix. [SRR-CWDA-2014-00080]

The NRC has revised the Monitoring Plan for FTF to include both FTF and HTF. The revised plan addresses the key monitoring areas for both tank farms. [ML15238A761] In the context of testing and research activities related to the FTF and HTF, the revised Monitoring Plan includes recommendations from the initial NRC FTF Monitoring Plan (ML12345A322) and any NRC TRRs issued subsequent to the NRC Monitoring Plan (see Appendix C). These recommendations will require further evaluation to determine how and when they should be addressed. The NRC recommendations are provided in Appendix C of this LW PA Maintenance Program Implementation Plan in the context of testing and research activities related to the FTF and HTF. In addition, Appendix C also contains NRC recommendations captured from NRC TRRs.

3.3.1 Tank Residual Characterization These tasks involve measurements and methods that will improve upon current knowledge of materials remaining in the waste tanks at operational closure. Some maintenance activities established under the SDF maintenance program (Section 2.0) may also inform the FTF PA and HTF PA such as those concerning cementitious degradation, soil parameters, and fracture formation.

3.3.1.1 Waste Release Studies Description: Through the NDAA Section 3116(a) consultation process, the NRC observed that uncertainties associated with the FTF PA doses might prevent DOE from meeting the 10 CFR Part 61, Subpart C performance objectives, particularly with regard to plutonium-related modeling assumptions. The NRC staff’s primary concern was that the timing of the FTF PA peak dose could be shifted into their period of performance (10,000 years) if certain assumptions were incorrect. This peak dose is principally associated with the residual Pu-239 inventory in Tank 18. The NRC’s TER recommends that DOE provide additional model support to further reduce the uncertainty surrounding PA assumptions that, if found to be significantly non-conservative, could result in this peak dose shifting into a 10,000-year performance period. [ML112371715]

FY2013: An experimental plan was developed in FY2013 to provide additional information regarding the residual waste solubility assumptions used in the FTF and HTF PA waste release models. This task was to be performed in two parts, the first part being development of the test plan and methods and the second part being conducting the actual waste testing with simulants and tank samples. The first part was completed in FY2013. [SRNL-RP-2013-00203]

FY2014: The overall objective of the task is to provide additional information regarding the residual waste solubility assumptions used in the FTF and HTF PA waste release models by developing a series of analytic methods to be used to test the solubility of plutonium, neptunium, uranium, and technetium under various simulated waste tank chemistry conditions using actual waste tank residuals. Waste release testing using simulants was initiated and the results were documented in Determining the Release of Radionuclides from Tank Waste Residual Solids. [SRNL-STI-2014-00456] Based on their findings, the authors recommend that leachate testing with surrogate materials be continued and finalized prior to initiating


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testing with actual radioactive materials. Testing with actual waste tank solids would be expected to provide additional information regarding the residual waste solubility assumptions used in the FTF and HTF PA waste release models.

FY2015: Waste release testing was performed in FY2015 in the areas of pore water development (including testing to understand and control stabilities of Oxidized Region II and Oxidized Region III) and surrogates solid testing (zero-head space and open-head space with O2 and CO2 present for Oxidized Region II and Oxidized Region III). The results of this testing was documented in an FY2015 testing report (Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2015 Report, SRNL-STI-2015-00446) in anticipation of actual waste testing in FY2016. FY2016: Testing of actual waste (i.e., Tank 18 residuals) was performed in FY2016 using the methodologies developed to date. The solubilities of Pu, Np, U, and Tc were tested under simulated waste tank chemistry conditions using Tank 18 residual waste samples, with the results documented in Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2016 Report (SRNL-STI-2016-00432).

A test apparatus was designed and constructed to simultaneously maintain numerous actual radioactive samples under either oxidizing or reducing conditions with continuous agitation and gas purge and at constant temperature. A photograph of the test apparatus prior to transfer into the SRNL shielded cells is provided in Figure 3.3-1. [SRNL-STI-2016-00432]

Data was collected via measured concentrations (or solubilities) over a seven-week testing period under different chemical conditions (pH and Eh were varied). The chemical conditions reflect a range of states such that the test results can be used to better understand the impact of transitory waste tank chemical conditions on solubility. As discussed in Evaluation of Waste Release Testing Results against the Tank Farm Performance Assessment Waste Release Model (SRR-CWDA-2016-00086), the experimental results indicate there may be some variance from the actual waste solubilities and the Waste Release Model (WRM) assigned solubilities. For example, Np was in all cases more insoluble than assigned in the WRM. The other three elements (Pu, Tc and U) appeared in most instances to be potentially more soluble than was assumed in the WRM. For example, Pu was relatively insoluble when oxidized, but was still more soluble than calculated by the WRM. Even if the experimental results were conservatively accepted into the WRM, the newly assigned solubilities would have a negligible impact on peak doses in 1,000 or 10,000 years. Because replacing the WRM values associated with Pu, Tc, U and Np does not significantly impact FTF and HTF PA conclusions, there is no immediate need to update the current model to incorporate the waste release testing data. The new waste release data can be integrated into the next revision to the FTF and HTF PAs. It should be noted that this waste release testing is distinct from the Lysimeter Program soil Kd testing discussed in Section 2.3.2.


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Figure 3.3-1: Photograph of Testing Equipment Prior to Installation into the SRNL Shielded Cell


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FY2017: Based on the successful testing of Tank 18 residuals in FY2016, the decision was made to test the residual solids from Tank 12. Tank 12 was an acid-cleaned tank versus the mechanical-only cleaning conducted on Tank 18 thus providing additional waste release data for comparison. In addition, Tank 12 is the tank which drives currently calculated doses in H-Tank Farm modeling. Since I-129 is the radionuclide that drives the calculated dose peak, iodine will also be analyzed in addition to Pu, Np, U and Tc as done for Tank 18. In FY2017 the Technical Task Request was generated by SRR, Waste Release Testing Program (G-TTR-H-00014), and the Task Technical and Quality Assurance Plan (TTQAP) was generated by SRNL, Task Technical and Quality Assurance Plan for Tank 12 Waste Residual Radionuclide Release Testing (SRNL-RP-2017-00411). The scope defined by the TTQAP in FY2017 included procurement of test equipment and initial chemical preparation of the Tank 12 solids. New equipment procurement was necessary as the equipment used for Tank 18 testing had been discarded during the SRNL High-Level Caves renovations. FY2018: The remaining scope of the TTQAP is anticipated to be completed in FY2018. The final equipment assembly, placement of equipment in the High-Level Caves, testing of Tank 12 residuals and final report issuance in planned in FY2018.

Deliverable: Test Plan (SRNL-RP-2013-00203) – Complete Residual Solids Technical Report (SRNL-STI-2014-00456) – Complete Residual Solids Technical Report (SRNL-STI-2015-00446) – Complete Waste Testing Technical Report (SRNL-STI-2016-00432) – Complete Waste Testing Evaluation (SRR-CWDA-2016-00086) – Complete Tank 12 Test Plan (SRNL-RP-2017-00411) - Complete

Expected Completion Date: FY2013 (Test Plan) – Complete FY2014 (Residual Solids Report) – Complete FY2015 (Residual Solids Report) – Complete FY2016 (Waste Testing Reports) – Complete FY2017 (Test Plan) – Complete FY2018 (Residual Solids Report) – FY2018



3.3.2 To Be Determined Out-Year Testing Description: For FY2018 and beyond, testing has not been finalized; however, for budget purpose estimated costs are included. Out-year testing may include projects such as waste release testing (similar to the testing discussed in Section 3.3.1.1) on actual waste from tanks other than Tank 18.




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Estimated Cost: FY2019 $540K, FY2020 $530K, FY2021 through FY2022 TBD in FY2019 or later


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4.0 REFERENCES

Note: References identified as (Copyright) were used in the development of this document, but are protected by copyright laws. No part of the publication may be reproduced in any form or by any means, including photocopying or electronic transmittal in any form by any means, without permission in writing from the copyright owner.

10 CFR 61, Licensing Requirements for Land Disposal of Radioactive Waste, U.S. Nuclear Regulatory Commission, Washington, DC, January 1, 2010.

ASTM C1260-14, Standard Test Method for Potential Alkali Reactivity of Aggregates, ASTM International, West Conshohocken, PA, 2007.

ASTM D5084-10, (COPYRIGHT) Standard Test Methods for Measurement of Hydraulic Conductivity of Saturated Porous Materials Using a Flexible Wall Permeameter, ASTM International, West Conshohocken, PA, 2010.

CBU-PIT-2005-00146, Saltstone Performance Objective Demonstration Document, Savannah River Site, Aiken, SC, Rev. 0, June 2005.

DOE M 435.1-1, Chg. 2, Radioactive Waste Management Manual, U.S. Department of Energy, Washington, DC, June 28, 2011.

DOE O 435.1, Chg. 1, Radioactive Waste Management, U.S. Department of Energy, Washington, DC, August 28, 2001.

DOE/SRS-WD-2012-001, Basis for Section 3116 Determination for Closure of F-Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, March 27, 2012.

DOE/SRS-WD-2013-001, Draft Basis for Section 3116 Determination for Closure of H-Tank Farm at Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, February 2013.

DOE/SRS-WD-2014-001, Basis for Section 3116 Determination for Closure of H-Tank Farm at Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, December 2014.

DOE_01-17-2006, Section 3116 Determination for Salt Waste Disposal at the Savannah River Site, Secretary of Energy, Washington, DC, January 17, 2006. DOE_03-28-2012, [Letter, Huizenga to Moody] Tier 1 Authorization to Proceed with Closure Activities in Accordance with Department of Energy Manual 435.1-1 for the F-Area Waste Tank Systems, U.S. Department of Energy, Washington, DC, March 28, 2012. DOE-OS-2015-04-27-01, [Letter, Whitney to Moody] Tier I Authorization to Proceed with Closure Activities in Accordance with DOE Manual 435.1-1, Radioactive Waste Management, for the H-Area Tank Farm, U.S. Department of Energy, Washington, DC, April 27, 2015.

DOE-STD-5002-2017, Disposal Authorization Statement and Tank Closure Documentation, U.S. Department of Energy, Washington, DC, May 2017.

DOE-WD-2005-001, Basis for Section 3116 Determination for Salt Waste Disposal at the Savannah River Site, Savannah River Site, Aiken, SC, January 2006.


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DOE-WD-2012-001, Section 3116 Determination for Closure of F-Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, March 2012.

DOE-WD-2014-001, Section 3116 Determination for Closure of H-Tank Farm at the Savannah River Site, U.S. Department of Energy, Washington, DC, December 2014.

EPA_Method_1315, Mass Transfer Rates of Constituents in Monolithic or Compacted Granular Materials Using a Semi-Dynamic Tank Leaching Procedure, U.S. Environmental Protection Agency, Washington, DC, January 2013.

G-TTR-H-00014, LW Technical Task Request - Waste Release Testing Program, Savannah River Site, Aiken, SC, June 2017.

LFRG_08-13-2008, Review Team Report for the Performance Assessment for the F-Tank Farm at the Savannah River Site, U.S. Department of Energy, Washington, DC, August 13, 2008.

ML112371715, Technical Evaluation Report for F-Area Tank Farm Facility, Savannah River Site, South Carolina, U.S. Nuclear Regulatory Commission, Washington, DC, October 27, 2011. ML12212A192, U.S. Nuclear Regulatory Commission Plan for Monitoring Disposal Actions Taken by the U.S. Department of Energy at the Savannah River Site F-Area Tank Farm Facility in Accordance With the National Defense Authorization Act for Fiscal Year 2005, U.S. Nuclear Regulatory Commission, Washington, DC, January 2013.

ML12272A082, Technical Review: Waste Release and Solubility Related Documents Prepared by United States Department of Energy to Support Final Basis for Section 3116 Determination for the F-Area Tank Farm Facility at Savannah River Site. Project No. PROJ0734, U.S. Nuclear Regulatory Commission, Washington, DC, May 21, 2013.

ML12272A124, Technical Review of Environmental Monitoring and Site-Specific Distribution Coefficient Reports, U.S. Nuclear Regulatory Commission, Washington, DC, March 31, 2015.

ML12345A322, U.S. Nuclear Regulatory Commission Planned Monitoring Activities for F-Area Tank Farm at Savannah River Site, U.S. Nuclear Regulatory Commission, Washington, DC, January 23, 2013.

ML13080A401, Technical Review Updated Cost-Benefit Analysis for Removal of Additional Highly Radioactive Radionuclides From Tank 18, U.S. Nuclear Regulatory Commission, Washington, DC, March 21, 2013.

ML13085A291, Technical Review of Final Inventory Documentation for Tanks 5 and 6 at F-Tank Farm, Savannah River Site, (Project No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, September 30, 2013.

ML13100A113, U.S. Nuclear Regulatory Commission Plan for Monitoring Disposal Actions Taken by the U.S. Department of Energy at the Savannah River Site Saltstone Facility in Accordance With the National Defense Authorization Act for Fiscal Year 2005, U.S. Nuclear Regulatory Commission, Washington, DC, Rev. 1, September 2013.

ML13100A230, Technical Review of Tank 18 and 19 Special Analysis for the Performance Assessment for the F-Tank Farm at the Savannah River Site, SRR-CWDA-2010-00124, Rev. 0, February 2012 (Docket No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, April 10, 2013.


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ML13269A365, Technical Review: U.S. Department Of Energy Documentation Related to Tanks 18F and 19F Final Configurations with an Emphasis on Grouting from Recommendations and Testing to Final Specifications and Procedures (Project No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, October 30, 2013.

ML13273A299, Technical Review of Tanks 5 and 6 Special Analysis at F-Tank Farm Facility, Savannah River Site (Project No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, October 4, 2013.

ML13277A063, Technical Review: U.S. Department of Energy Documentation Related to Features, Events, and Processes in the F-Area Tank Farm Performance Assessment, U.S. Nuclear Regulatory Commission, Washington, DC, April 21, 2014.

ML13304B159, Technical Review: Solubility of Technetium Dioxides in Reducing Cementitious Material Leachates, A Thermodynamic Calculation, Docket No. PROJ0734, U.S. Nuclear Regulatory Commission, Washington, DC, November 7, 2013. ML14094A496, Technical Evaluation Report for Draft Waste Determination for Savannah River Site H Area Tank Farm, U.S. Nuclear Regulatory Commission, Washington, DC, June 17, 2014.

ML14342A784, Technical Review: U.S. Department of Energy Documentation Related to Tanks 5F and 6F Final Configurations with an Emphasis on Grouting From Recommendations and Testing to Final Specifications and Procedures (Project No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, December 16, 2014.

ML15098A031, Technical Review: Oxidation of Reducing Cementitious Waste Forms, Docket No. PROJ0734, U.S. Nuclear Regulatory Commission, Washington, DC, June 4, 2015.

ML15238A761, U.S. Nuclear Regulatory Commission Plan for Monitoring Disposal Actions Taken by the U.S. Department of Energy at the Savannah River Site F-Area and H-Area Tank Farm Facilities in Accordance with the National Defense Authorization Act for Fiscal Year 2005, U.S. Nuclear Regulatory Commission, Washington, DC, October 2015.

ML15301A710, Technical Review of “Tank 16H Special Analysis for the Performance Assessment for the H-Tank Farm at the Savannah River Site”, SRR-CWDA-2014-00106, Rev. 1, February 2015 (Project No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, November 10, 2015.

ML15301A830, Technical Review of Final Inventory Documentation for Tank 16H at Savannah River Site (Project No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, November 6, 2015.

ML16196A179, Technical Review of Quality Assurance Documentation for the Cementitious Barriers Partnership Toolbox (Project No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, August 1, 2016.

ML16231A444, Technical Review: U.S. Department of Energy Documentation Related to Tanks 16H and 12H Grouting Operations with Emphases on Specifications, Testing, Recommendations and Placement Procedures (Project No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, September 6, 2016.


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ML16277A060, Technical Review of “Dose Calculation Methodology for Liquid Waste Performance Assessments at the Savannah River Site”, SRR-CWDA-1013-00058, Rev. 1, July 2014, Docket No. PROJ0734, U.S. Nuclear Regulatory Commission, Washington, DC, December 23, 2016.

ML16342C575, Technical Review: Iodine Sorption Coefficients for Use in Performance Assessments for the Saltstone Disposal Facility, Docket No. PROJ0734, U.S. Nuclear Regulatory Commission, Washington, DC, January 5, 2017.

ML17018A137, Technical Review Report Saltstone Waste Form Hydraulic Performance (Docket No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, March 23, 2017.

ML17081A187, Performance of the High Density Polyethylene Layer, High Density Polyethylene/Geosynthetic Clay Liner Composite Layer, and the Lower Lateral Drainage Layer, Docket No. PROJ0734, U.S. Nuclear Regulatory Commission, Washington, DC, April 12, 2017.

ML17097A351, Closure of Monitoring Factors in the 2013 U.S. Nuclear Regulatory Commission Saltstone Disposal Facility Monitoring Plan (Docket No. PROJ0734), U.S. Nuclear Regulatory Commission, Washington, DC, June 5, 2017.

NDAA Section 3116, Public Law 108-375, Ronald W. Reagan National Defense Authorization Act for Fiscal Year 2005, Section 3116, Defense Site Acceleration Completion, October 28, 2004.

NUREG CR-7028, Benson, C.H., et al., Engineered Covers for Waste Containment: Changes in Engineering Properties and Implications for Long-Term Performance Assessment, U.S. Nuclear Regulatory Commission, Washington, DC, December 2011.

PNNL-21723, Cantrell, K. J., and Williams, B. D., Equilibrium Solubility Model for Technetium Release from Saltstone Based on Anoxic Single-Pass Flow through Experiments, Pacific Northwest National Laboratory, Richland, WA, September 2012.

SCDHEC R.61-67, Standards for Wastewater Facility Construction, South Carolina Department of Health and Environmental Control, Columbia, SC, May 22, 2015.

SCDHEC R.61-82, Proper Closeout of Wastewater Treatment Facilities, South Carolina Department of Health and Environmental Control, Columbia, SC, April 11, 1980.

SIMCO_08-31-2012, Final Report Comparison of Wasteform Mixtures, SIMCO Technologies, Inc., Quebec, QC, Canada, August 2012.

SREL Doc. R-13-0004, Seaman, J.C. and Chang, H., Impact of Cementitious Material Leachate on Contaminant Partitioning, Savannah River Site, Aiken, SC, Version 1.0, September 4, 2013.

SREL Doc. R-13-0005, Seaman, J.C. and Chang, H., Impact of Cementitious Leachate on Se, Nb and Ra Partitioning, Savannah River Site, Aiken, SC, Version 1.0, September 5, 2013.

SREL Doc. R-13-0006, Seaman, J.C., et al., Comparison of Hydraulic Property Measurement Techniques for Simulated Saltstone, Savannah River Site, Aiken, SC, Version 1.0, September 4, 2013.

SREL Doc. R-14-0005, Seaman, J.C. and Chang, H., Impact of Cementitious Material Leachate on Iodine Partitioning, Savannah River Site, Aiken, SC, Version 1.0, September 16, 2014.


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SREL Doc. R-14-0006, Seaman, J.C., et al., Chemical and Physical Properties of Saltstone as Impacted by Curing Duration, Savannah River Site, Aiken, SC, Version 1.0, September 23, 2014.

SREL Doc. R-14-0007, Seaman, J.C. and Chang, H., Dynamic Leaching Characterization of Saltstone, Savannah River Site, Aiken, SC, Version 1.0, September 29, 2014.

SREL Doc. R-15-0003, Seaman, J.C., Chemical and Physical Properties of 99Tc-Spiked Saltstone as Impacted by Curing Duration and Leaching Atmosphere, Savannah River Site, Aiken, SC, October 1, 2015.

SREL Doc. R-16-0003, Seaman, J.C., et al., Contaminant Leaching from Saltstone, Savannah River Site, Aiken, SC, September 29, 2016.

SREL Doc. R-17-0004, Seaman, J.C. and Thomas, R.J., Impact of Cementitious Material Leachate on Iodine Partitioning, Savannah River Site, Aiken, SC, Version 1.0, September 29, 2017.

SREL Doc. R-17-0005, Seaman, J.C. and Coutelot, F.M., Contaminant Leaching from Saltstone, Savannah River Site, Aiken, SC, September 29, 2017. SRNL-L3100-2015-00073, Experimental Plan for SRNL Support for Saltstone Sampling and Analysis, Rev. 2, Savannah River Site, Aiken, SC, January 2016.

SRNL-RP-2013-00203, Hobbs, D.T. et. al, Task Technical and Quality Assurance Plan for Determining the Radionuclide Release from Tank Waste Residual Solids, Savannah River Site, Aiken, SC, Rev. 3, October 7, 2015.

SRNL-RP-2017-00411, King, W.D., Task Technical and Quality Assurance Plan for Tank 12 Waste Residual Radionuclide Release Testing, Savannah River Site, Aiken, SC, Rev. 0, August 2017.

SRNL-STI-2008-00421, Dixon, K., et al., Hydraulic and Physical Properties of Saltstone Grouts and Vault Concretes, Savannah River Site, Aiken, SC, November 2008.

SRNL-STI-2009-00419, Dixon, K., et al., Hydraulic and Physical Properties of ARP/MCU Saltstone Grout, Savannah River Site, Aiken, SC, May 2010.

SRNL-STI-2009-00473, Kaplan, D.I., Geochemical Data Package for Performance Assessment Calculations Related to the Savannah River Site, Savannah River Site, Aiken, SC, March 15, 2010.

SRNL-STI-2009-00512, Savannah River Site DOE 435.1 Composite Analysis, Volumes I and II, Savannah River Site, Aiken, SC, Rev. 0, June 10, 2010.

SRNL-STI-2010-00167, Design and Testing of the Formed-Core Sampling System for Saltstone Facility Vaults, Savannah River Site, Aiken, SC, Rev. 0, March 15, 2010.

SRNL-STI-2011-00551, (SUPERSEDED) Stefanko, D.B. and C.A. Langton, Tanks 18 and 19-F Structural Flowable Grout Fill Material Evaluation and Recommendations, Savannah River Site, Aiken, SC, Rev. 0, September 2011.

SRNL-STI-2011-00661, Dixon, K., Moisture Retention Properties of High Temperature Cure ARP/MCU Saltstone Grout, Savannah River Site, Aiken, SC, Rev. 0, December 2011.


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SRNL-STI-2011-00665, Reigel, M.M., et al., Operational and Compositional Factors that Affect the Performance Properties of ARP/MCU Saltstone Grout, Savannah River Site, Aiken, SC, Rev. 0, February 2012.

SRNL-STI-2011-00672, Almond, P.M., et al., Variability of Kd Values in Cementitious Materials and Sediments, Savannah River Site, Aiken, SC, Rev. 0, January 2012.

SRNL-STI-2011-00716, Estes, S.L., et al., Technetium Sorption of Cementitious Materials Under Reducing Conditions, Savannah River Site, Aiken, SC, Rev. 0, January 2012.

SRNL-STI-2012-00087, Denham, M., Evolution of Chemical Conditions and Estimated Plutonium Solubility in the Residual Waste Layer During Post-Closure Aging of Tank 18, Savannah River Site, Aiken, SC, Rev. 0, February 2012.

SRNL-STI-2012-00138, Kaplan, D.I., Revised Guidelines for using Cellulose Degradation Product-Impacted Kd Values for Performance Assessments and Composite Analyses, Savannah River Site, Aiken, SC, Rev. 0, May 2012. SRNL-STI-2012-00267, Langton, C.A., Transport Through Cracked Concrete: Literature Review, Savannah River Site, Aiken, SC, Rev. 0, April 2012.

SRNL-STI-2012-00435, Turrick, C.E. and Berry, C.J., Review of Concrete Biodeterioration in Relation to Buried Nuclear Waste, Savannah River Site, Aiken, SC, Rev. 0, August 2012.

SRNL-STI-2012-00468, Almond, P.M., et al., Method Evaluation and Field Sample Measurements for the Rate of Movement of the Oxidation Front in Saltstone, Savannah River Site, Aiken, SC, Rev. 0, August 2012.

SRNL-STI-2012-00551, Miller, D.H. and Reigel, M.M., Formed Core Sampler Hydraulic Conductivity Testing, Savannah River Site, Aiken, SC, Rev. 0, September 2012.

SRNL-STI-2012-00558, Reigel, M.M., et al., Process Formulations and Curing Conditions that Affect Saltstone Properties, Savannah River Site, Aiken, SC, Rev. 0, September 2012.

SRNL-STI-2012-00596, Estes, S.L., et al., Technetium Sorption by Cementitious Materials Under Reducing Conditions, Savannah River Site, Aiken, SC, Rev. 0, September 2012.

SRNL-STI-2012-00603, Bagwell, L., et al., SRNL Radionuclide Field Lysimeter Experiment Baseline Construction and Implementation, Savannah River Site, Aiken, SC, Rev. 0, October 18, 2012.

SRNL-STI-2013-00522, Dixon, K.L. and Nichols, R.L., Method Development for Determining the Hydraulic Conductivity of Fractured Porous Media, Savannah River Site, Aiken, SC, Rev. 0, September 2013.

SRNL-STI-2013-00533, Nichols, R.L. and Dixon, K.L., Saltstone Osmotic Pressure, Savannah River Site, Aiken, SC, September 2013.

SRNL-STI-2014-00456, Miller, D.H., et al., Determining the Release of Radionuclides from Tank Waste Residual Solids, Savannah River Site, Aiken, SC, Rev. 0, September 2014.

SRNL-STI-2015-00446, King, W.D. and Hobbs, D.T., Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2015 Report, Savannah River Site, Aiken, SC, Rev. 0, September 2015.


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SRNL-STI-2016-00106, Results and Analysis of Saltstone Cores Taken from Saltstone Disposal Unit Cell 2A, Rev. 0, March 2016.

SRNL-STI-2016-00432, Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2016 Report, Savannah River Site, Aiken, SC, Rev. 0, August 2016.

SRNS-STI-2008-00045, Kaplan, D.I., et al., Saltstone and Concrete Interactions with Radionuclides: Sorption (Kd), Desorption, and Reduction Capacity Measurements, Savannah River Site, Aiken, SC, October 30, 2008.

SRRA021685, Witmer, M. and Powell, B.A., Determination of Constituent Concentrations in Field Lysimeter Effluents FY14 Final Report, Clemson University, Clemson, SC, Rev. 0, September 25, 2014.

SRRA021685-000007, Pope, R. et al., Determination of Constituent Concentrations in Field Lysimeter Effluents, Clemson University, Clemson, SC, September 30, 2015.

SRRA021685-000008, Powell, B.A., et al., Determination of Constituent Concentrations in Field Lysimeter Effluents, Clemson University, Clemson, SC, September 28, 2017. SRRA021685-000009, Powell, B.A., et al., Analysis of Plutonium Soil Concentrations in Field Lysimeter Experiments, Clemson University, Clemson, SC, September 28, 2017.

SRRA042328-000004, Arai, Y. and Powell, B.A., Examination of Tc, S, and Fe Speciation within Saltstone, Clemson University, Clemson, SC, September 25, 2015.

SRRA042328SR, Arai, Y. and Powell, B.A., Examination of Tc, S, and Fe Speciation within Saltstone, Clemson University, Clemson, SC, Rev. 0, September 30, 2014.

SRR-CWDA-2009-00017, Performance Assessment for the Saltstone Disposal Facility at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, October 29, 2009.

SRR-CWDA-2009-00054, Comment Response Matrix for NRC Comments on Revision 0 of SRS-REG-2007-00002, Performance Assessment for F-Tank Farm, Savannah River Site, Aiken, SC, Rev. 0, February 19, 2010.

SRR-CWDA-2010-00015, Savannah River Site Liquid Waste Facilities Performance Assessment Maintenance Program, FY2010 Implementation Plan, Savannah River Site, Aiken, SC, Rev. 0, March 31, 2010.

SRR-CWDA-2010-00124, Tank 18/Tank 19 Special Analysis for the Performance Assessment for the F-Area Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, February 2012.

SRR-CWDA-2010-00128, Performance Assessment for the H-Area Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, March 14, 2010.

SRR-CWDA-2010-00147, Tier 1 Closure Plan for the F-Area Waste Tank Systems at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, March 20, 2012.

SRR-CWDA-2011-00015, F-Area Tank Farm Tanks 18 and 19 Tier 2 Closure Plan, Savannah River Site, Aiken, SC, Rev. 0, March 22, 2012.


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SRR-CWDA-2011-00054, Comment Response Matrix for United States Nuclear Regulatory Commission Staff Comments on the Draft Basis for Section 3116 Determination and Associated Performance Assessment for the F-Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, October 25, 2011.

SRR-CWDA-2011-00135, Responses to South Carolina Department of Health & Environmental Control Comments on Performance Assessment for the H-Area Tank Farm at Savannah River Site (SRR-CWDA-2010-00128, Rev. 0 March 2011) Received March 24, 2011, Savannah River Site, Aiken, SC, Rev. 1, February 6, 2012.

SRR-CWDA-2011-00164, Comment Response Matrix for South Carolina Department of Health and Environmental Control (SCDHEC) Comments on: Performance Assessment for the F-Tank Farm, SRS-REG-2007-00002, Revision 1, March 31, 2010, Savannah River Site, Aiken, SC, Rev. 0, February 2012.

SRR-CWDA-2011-00175, Comment Response Matrix for United States Environmental Protection Agency Comments on: Performance Assessment for the F-Tank Farm, SRS-REG-2007-00002, Revision 1, March 31, 2010 and Comment Response Matrix for United States Environmental Protection Agency Comments on Revision 0 of the F-Tank Farm Performance Assessment SRR-CWDA-2009-00055, Revision 0, February 16, 2010, Savannah River Site, Aiken, SC, Rev. 0, February 2012.

SRR-CWDA-2012-00022, Evaluation of the Features, Events, and Processes in the F-Area Tank Farm Performance Assessment, Savannah River Site, Aiken, SC, Rev. 0, February 7, 2012.

SRR-CWDA-2012-00070, Performance Assessment for the H-Area Tank Farm at the Savannah River Site: Quality Assurance Report, Savannah River Site, Aiken, SC, Rev. 0, August 30, 2012.

SRR-CWDA-2012-00106, Tanks 5 and 6 Special Analysis for the Performance Assessment for the F-Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 1, January 2013.

SRR-CWDA-2012-00166, Responses to EPA Comments on “Responses to EPA Comments On Performance Assessment for the H-Area Tank Farm At the Savannah River Site SRR-CWDA-2010-00128, Rev 0 March 2011”, Savannah River Site, Aiken, SC, Rev. 0, November 2012.

SRR-CWDA-2012-00170, Tanks 18 and 19 Final Configuration Report for F-Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, June 2013.

SRR-CWDA-2013-00014, F-Area Tank Farm Tanks 5 and 6 Tier 2 Closure Plan, Savannah River Site, Aiken, SC, Rev. 0, May 2013.

SRR-CWDA-2013-00026, Performance Assessment Monitoring Plan for the Saltstone Disposal Facility at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 1, August 17, 2015.

SRR-CWDA-2013-00037, Closure Plan for the Z-Area Saltstone Disposal Facility, Savannah River Site, Aiken, SC, Rev. 1, September 2, 2015.

SRR-CWDA-2013-00058, Dose Calculation Methodology for Liquid Waste Performance Assessments at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 1, July 2014

SRR-CWDA-2013-00062, FY2013 Special Analysis for the Saltstone Disposal Facility at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 2, October 2013.


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SRR-CWDA-2013-00121, Witmer, M. and Powell, B.A., Determination of Constituent Concentrations in Field Lysimeter Effluents, FY13 Final Report – Task 2: Lysimeter Leachate Chemistry, Clemson University, Clemson, SC, Rev. 0, September 6, 2013.

SRR-CWDA-2013-00123, Hatfield A., et al., Examination of Cr and S Speciation within Saltstone Monoliths, FY13 Final Report – Task 1: Saltstone Redox Chemistry, Clemson University, Clemson, SC, Rev. 0, September 11, 2013.

SRR-CWDA-2014-00006, FY2014 Special Analysis for the Saltstone Disposal Facility at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 2, September 2014.

SRR-CWDA-2014-00040, Tier 1 Closure Plan for the H-Area Waste Tank Systems at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, December 2014.

SRR-CWDA-2014-00059, FY2014 Saltstone Core-Drilling Mock-Up Summary, Savannah River Site, Aiken, SC, Rev. 0, June 2014.

SRR-CWDA-2014-00060, Updates to the H-Area Tank Farm Stochastic Fate and Transport Model, Savannah River Site, Aiken, SC, Rev. 2, June 2016. SRR-CWDA-2014-00080, Nuclear Regulatory Commission’s H-Tank Farm Technical Evaluation Reports Recommendations - Department of Energy’s Activity Summary Matrix, Aiken, SC, Rev. 0, December 2014.

SRR-CWDA-2014-00106, Tank 16 Special Analysis for the Performance Assessment for the H-Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 1, February 2015.

SRR-CWDA-2014-00108, Savannah River Site Liquid Waste Facilities Performance Assessment Maintenance Program FY2015 Implementation Plan, Savannah River Site, Aiken, SC, Rev. 0, January 2015.

SRR-CWDA-2015-00002, FY2015 Saltstone Core-Drilling Mock-Up Summary, Savannah River Site, Aiken, SC, Rev. 0, February 2015.

SRR-CWDA-2015-00009, H-Tank Farm Tank 16 Tier 2 Closure Plan, Savannah River Site, Aiken, SC, Rev. 0, May 2015.

SRR-CWDA-2015-00119, H-Tank Farm Tank 12 Tier 2 Closure Plan, Savannah River Site, Aiken, SC, Rev. 0, September 2015.

SRR-CWDA-2015-00158, H-Tank Farm Type I and Type II Tank Special Analysis Base Case Model Inputs, Savannah River Site, Aiken, SC, Rev. 1, March 4, 2016.

SRR-CWDA-2015-00066, Summary of Saltstone Disposal Unit Cell 2A Core Drill Activities, Savannah River Site, Aiken, SC, Rev. 0, May 2015.

SRR-CWDA-2015-00073, Tank 12 Special Analysis for the Performance Assessment for the H-Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, August 2015.

SRR-CWDA-2016-00051, Property Data for Core Samples Extracted from SDU Cell 2A, Savannah River Site, Aiken, SC, Rev. 0, April 2016.

SRR-CWDA-2016-00072, FY2016 Special Analysis for the Saltstone Disposal Facility at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, October 2016.


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SRR-CWDA-2016-00078, Type I and II Tanks Special Analysis for the Performance Assessment for the H-Tank Farm at the Savannah River Site, Savannah River Site, Aiken, SC, Rev. 0, August 2016.

SRR-CWDA-2016-00086, Evaluation of Waste Release Testing Results Against the Tank Farm Performance Assessment Waste Release Model, Savannah River Site, Aiken, SC, Rev. 0, August 2016.

SRR-SPT-2012-00049, Saltstone Sampling and Analyses Plan, Savannah River Site, Aiken, SC, Rev. 1, May 23, 2013.

SRR-SPT-2013-00044, Update of Fiscal Year 2013 Activities Related to SDU Sampling and Analyses, Savannah River Site, Aiken, SC, Rev. 0, September 6, 2013.

SRS-REG-2007-00002, Performance Assessment for F-Tank Farm at the Savannah River Site Savannah River Site, Aiken, SC, Rev. 1, March 31, 2010.

U-ESR-H-00107, Clark, D.J., Tank 16H: Preliminary Evaluation of Cessation of Annulus Waste Removal Activities, Savannah River Site, Aiken, SC, Rev. 0, March 2013. VSL-13R3010-1, Papathanassiu, A.E., et al., Oxidation Rate Measurement and Humidity Effects on Saltstone, Vitreous State Laboratory, The Catholic University of America, Washington, DC 20064, August 22, 2013.

VSL-14R3210-1, Papathanassiu, A.E., et al., Final Report - Saltstone Curing Assessment, Vitreous State Laboratory, The Catholic University of America, Washington, DC 20064, Rev. 0, March 27, 2014.

WDPD-12-39, [Letter, Langton to Moody] Request for Contracting Officer Approval – Closure Documents Required for Operational Closure of Tanks 18 and 19, U.S. Department of Energy, Savannah River Site, Aiken, SC, March 29, 2012.

WDPD-12-49, [Letter, Folk to Olsen] Disposal Authorization Statement for the Savannah River Site Saltstone Disposal Facility, U.S. Department of Energy, Savannah River Site, Aiken, SC, May 22, 2012.

WDPD-13-56, [Letter, Moody to Schlismann] Contract DE-AC09-09SR22505 - Request for Contracting Officer Approval - Closure Documents Required for Operational Closure of Tanks 5 and 6 (Ltr, Schlismann to Smith, SRR-CAA-2013-00160, 5/13/13), U.S. Department of Energy, Savannah River Site, Aiken, SC, May 17, 2012.

WDPD-14-08, [Letter, Spears to Camper], Transmittal of Fiscal Year 2013 Special Analysis for the Saltstone Disposal Facility at the Savannah River Site SRR-CWDA-2013-00062, Rev. 2, U.S. Department of Energy, Aiken, SC, December 5, 2013.

WDPD-15-42, [Letter, Moody to Bair] Contract DE-AC09-09SR22505 - Request for Contracting Officer Approval - Closure Documents Required for Operational Closure of Tank 16 (Ltr, Bair to Smith, SRR-CAA-2015-00141, 5/4/15), U.S. Department of Energy, Savannah River Site, Aiken, SC, May 17, 2012.

WDPD-16-17, [Letter, Smith to Bair], Contract DE-AC09-09SR22505 – Request for Contracting Officer Approval Closure Documents Required for Operational Closure of Tank 12H, U.S. Department of Energy, Savannah River Site, Aiken, SC, December 2015.


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WDPD-17-05, [Letter, Ridley to Fortenberry], Contract DE-AC09-09SR22505 – Department of Energy (DOE) Approval of the FY2016 Special Analysis for the Saltstone Disposal Facility at the Savannah River Site (SRR-CWDA-2016-00072), Revision 0, October 2016), U.S. Department of Energy, Savannah River Site, Aiken, SC, November 21, 2016.

WDPD-17-34, [Letter, Folk to Foster], Contract DE-AC09-09SR22505 – Department of Energy (DOE) Approval of the Type I and II Tanks Special Analysis for the Performance Assessment for the H-Tank Farm (HTF) at the Savannah River Site (SRR-CWDA-2016-00078, Revision 0, August 2016), U.S. Department of Energy, Savannah River Site, Aiken, SC, June 27, 2017.

WSRC-OS-94-42, Federal Facility Agreement for the Savannah River Site, Savannah River Site, Aiken, SC, August 16, 1991.

WSRC-RP-92-1360, Radiological Performance Assessment for the Z-Area Saltstone Disposal Facility, Savannah River Site, Aiken, SC, Rev. 0, December 1992.

WSRC-STI-2008-00244, Jones, W.E. and Phifer, M.A., Saltstone Disposal Facility Closure Cap Concept and Infiltration Estimates, Savannah River Site, Aiken, SC, Rev. 0, May 2008. WSRC-TR-2005-00074, Cook, J.R., et al., Special Analysis: Revision of Saltstone Vault 4 Disposal Limits, Savannah River Site, Aiken, SC, Rev. 0, May 26, 2005.


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APPENDIX A

Summary Tables for the Liquid Waste Facilities Performance Assessment Maintenance Program FY2018 Implementation Plan


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A.1 Summary Tables for the LW PA Maintenance Program Tables A.1-1 and A.1-2 summarizes the estimated expenditures by activity and FY. Table A.1-3 contains a summary of the combined estimated expenditures for all the LW facility PA maintenance activities. This Implementation Plan reflects the PA related activities in the annual operating plan for the current FY and the projected out-year activities for estimation purposes.


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Table A.1-1: Summary for the Saltstone Disposal Facility Performance Assessment Maintenance Program ($K)

Section Maintenance Activity FY18 FY19 FY20 FY21 FY22 Tasks Performed Annually

2.1.1 Maintain SDF PA Control Through UWMQ Process 85 85 85 85 85

2.1.2 Prepare Annual PA Maintenance Program Implementation Plan 15 15 15 15 15

2.1.3 Provide General Technical Support on SDF PA Issues 575 575 575 575 575

2.1.4 Develop and Maintain PA Model Revision Archive and Revision Control 5 5 5 5 5

2.1.5 Conduct Annual SDF PA Validation 15 15 15 15 15 2.1.6 Maintain SDF Closure Plan 5 5 5 5 5 2.1.7 Maintain SDF PA Monitoring Plan 5 5 5 5 5

Annual Tasks Total 705 705 705 705 705 Performance Assessment Development/Revisions

2.2.1 Prepare Out-Year SDF PA Revisions 500 350 0 0 0 2.2.2 SDF Special Analyses 0 0 0 0 0

PA Development/Revisions Total 500 350 0 0 0 Testing and Research Activities

2.3.1.1 Measurement of Kd in SRS Subsurface Sediments 0 0 0 0 0

2.3.1.2 Contaminant Leaching Characteristics from Saltstone Monolith 0 0 0 0 0

2.3.2.1 Long-Term Radiological Lysimeter Program 41 0 0 0 0

2.3.2.2 Studies Related to Concrete Degradation Due to Radiation Damage 0 0 0 0 0

2.3.2.3 Closure Cap Drainage Layer Long-Term Performance 65 0 0 0 0

2.3.2.4 Hydraulic Degradation Mechanisms in SDU Concrete 65 0 0 0 0

2.3.3.1

Measure Physical Properties of Laboratory Prepared Saltstone Simulant Samples, Actual Tank 50 Salt Solution Samples, Saltstone In-Line Process Sample, and SDU Cell 2A Emplaced Core Sample

130 0 0 0 0

2.3.4 To Be Determined Out-Year Testing N/A 655 530 TBD TBD Testing and Research Total 301 655 530 TBD TBD

SDF PA COMPILED TOTAL 1,506 1,710 1,235 705 705


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Table A.1-2: Summary for the Tank Farm Performance Assessment Maintenance Program ($K)

Section Maintenance Activity FY18 FY19 FY20 FY21 FY22

Tasks Performed Annually

3.1.1 Maintain Tank Farm PA Control Through UWMQ Process 170 170 170 170 170

3.1.2 Prepare Annual PA Maintenance Program Implementation Plan 30 30 30 30 30

3.1.3 Provide General Technical Support on Tank Farm PA Issues 370 370 370 370 370

3.1.4 Develop and Maintain PA Model Revision Archive and Revision Control 10 10 10 10 10

Annual Tasks Total 580 580 580 580 580

Performance Assessment Development/Revisions

3.2.1 Prepare Out-Year FTF PA Revisions 0 0 0 0 0 3.2.2 Prepare Out-Year HTF PA Revisions 0 0 500 500 0 3.2.3 Tank Farm Special Analyses 0 0 0 0 0

PA Development/Revisions Total 0 0 500 500 0

Testing and Research Activities

2.3.2.1 Long-Term Radiological Lysimeter Program 75 0 0 0 0 3.3.1.1 Waste Release Studies 565 0 0 0 0 3.3.2 To Be Determined Out-Year Testing N/A 540 530 TBD TBD

Testing and Research Total 640 540 530 TBD TBD

Tank Farm PA COMPILED TOTAL 1,220 1,120 1,610 1,080 580


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Table A.1-3: Summary for the Liquid Waste Performance Assessment Maintenance Program ($K)

LW PA Maintenance Program FY18 FY19 FY20 FY21 FY22

SDF PA Maintenance Program Totals 1,506 1,710 1,235 705 705

Tank Farm PA Maintenance Program Totals 1,220 1,120 1,610 1,080 580 COMPILED TOTAL 2,726 2,830 2,845 1,785 1,285


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APPENDIX B

U.S. Nuclear Regulatory Commission Monitoring Factors for Saltstone Disposal Facility


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Appendix B: U.S. Nuclear Regulatory Commission Monitoring Items for the Saltstone Disposal Facility Factors are colored by NRC priority and a symbol is included with each Monitoring Factor (MF) number to ensure clarity. A legend containing a description of the NRC ranking is provided at the end of the table. [ML13100A113]

MF #1 Factor NRC Expectations2 Related PA Maintenance Activities as Described in this Document Comments

Monitoring Activity (MA 1) – Inventory

1.01 §

Inventory in Disposal Structures

NRC expects to close MF 1.01 after DOE has completed waste disposal at the SDF and determined the final inventory in each disposal structure.

Section 2.1.5 - Conduct Saltstone Disposal Facility Performance Assessment Validation Contaminant inventories are determined based on sample analysis and to ensure compliance with the Saltstone waste acceptance criteria (WAC). The SDF inventory will be updated based on the results of quarterly sample results and transfer volumes. Annual review reports including updated actual inventory will be provided to the NRC.

DOE will continue to revise inventory estimates as part of waste disposal activities.

1.02 ‡

Methods Used to Assess Inventory

NRC expects to close MF 1.02 after DOE has completed waste disposal at the SDF and determined the final inventory in each disposal structure.

Section 2.1.5 - Conduct Saltstone Disposal Facility Performance Assessment Validation Contaminant inventories are determined based on sample analysis and to ensure compliance with the Saltstone WAC. The SDF inventory will be updated based on the results of quarterly sample results and transfer volumes. Annual review reports including updated actual inventory will be provided to the NRC.

Final inventory will be determined prior to SDF closure. PA Maintenance Plans indicate that a “future revision of the SDF PA will be scheduled as required.” This activity will include revised assessments of inventories to use in modeling. In FY2017 estimates of total Tc-99 and I-129 inventories were completed (SRR-CWDA-2015-00123, SRR-CWDA-2015-00077). Efforts began in late FY2017 to sample the supernate from 5 tanks estimated to contain a significant fraction of the total Tc and I inventories. After issuance of the analysis in FY2018, WDA will revise the previous reports to incorporate the new data.


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MA 2 – Infiltration and Erosion Control

2.01 †

Hydraulic Performance of Closure Cap

NRC expects to close MF 2.01 after NRC determines that the hydraulic performance of the as-built closure cap is adequate. Given the importance of construction activities on the performance of the cap, MF 2.01 will not be closed prior to construction of the cap.

Section 2.3.2.3 - Closure Cap Long-Term Performance Research will validate assumptions in the modeling concerning the rate of pluggage of the closure cap drainage layer as well as the drainage layer above each SDU.

Research began in FY2017 and will conclude in FY2018 to validate PA assumptions concerning the rate of closure cap infiltration as well as the drainage layer behavior. DOE will revise closure cap modeling assumptions and support once a final closure design has been determined.

2.02

†

Erosion Protection

NRC expects to close MF 2.02 after NRC determines that the physical stability of the final closure cap is adequate. Given the importance of construction activities on the performance of the cap, MF 2.02 will not be closed prior to construction of the cap.

N/A DOE will revise closure cap modeling assumptions and support once a final closure design has been determined.

MA 3 – Waste Form Hydraulic Performance

3.01 ±

Hydraulic Conductivity of Field-Emplaced Saltstone

NRC expects to close MF 3.01 after NRC determines that model support for the SHC of field-emplaced saltstone is sufficient.

Section 2.3.3.1 - Measure Physical Properties of Laboratory Prepared Saltstone Simulant Samples, Actual Tank 50 Salt Solution Samples, Saltstone In-Line Process Sample, and SDU Cell 2A Emplaced Core Sample Laboratory prepared and process room samples will have physical properties testing performed to determine the hydraulic conductivity, Kd, bulk cured density, porosity, and micro structure/phase analysis. Future testing will compare these properties to those measured from emplaced core sampling.

Saltstone testing of measured hydraulic conductivities is ongoing. A variety of laboratory testing has completed, including one in which samples were cured under conditions similar to those expected for field-emplaced saltstone. The results were incorporated into the FY2014 SDF SA. Multiple cores were extracted via a wet core drilling process in FY2015 approximately 20 months after the saltstone of interest was processed in the SPF and subsequently emplaced in SDU Cell 2A. The physical property data for SDU-emplaced and laboratory-prepared samples is summarized in the SDU Cell 2A Core Sampling Report (SRR-CWDA-2016-00051), with values for SDF model inputs provided in the report where applicable. Based on the these results, NRC closed this monitoring factor in June 2017 per ML17097A351.


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3.02 ±

Variability of Field-Emplaced Saltstone

NRC expects to close MF 3.02 after NRC determines that saltstone production, placement, and curing conditions that significantly affect saltstone hydraulic properties are well controlled.

Section 2.3.1 – Critial Property Testing Previous testing and research activities were carried out to define the operating conditions (e.g., water-to-premix ratio, dry feeds variability, and the curing temperature) required to meet or exceed PA expectations of saltstone performance. Section 2.3.3.1 - Measure Physical Properties of Laboratory Prepared Saltstone Simulant Samples, Actual Tank 50 Salt Solution Samples, Saltstone In-Line Process Sample, and SDU Cell 2A Emplaced Core Sample Laboratory prepared and process room samples will have physical properties testing performed to determine the hydraulic conductivity, Kd, bulk cured density, porosity, and micro structure/phase analysis. Future testing will compare these properties to those measured from emplaced core sampling.

Saltstone testing of measured hydraulic conductivities is ongoing. A variety of laboratory testing has completed, including one in which sampled were cured under conditions similar to those expected for field-emplaced saltstone. The results were incorporated into the FY2014 SDF SA. Variability of saltstone hydraulic conductivities was evaluated in the FY2014 SDF SA through the use of parametric flow cases that applied average, upper bounding, and lower bounding values.

Multiple cores were extracted via a wet core drilling process in FY2015 approximately 20 months after the saltstone of interest was processed in the SPF and subsequently emplaced in SDU Cell 2A. The physical property data for SDU-emplaced and laboratory-prepared samples is summarized in the SDU Cell 2A Core Sampling Report (SRR-CWDA-2016-00051), with values for SDF model inputs provided in the report where applicable. Based on the these results, NRC closed this monitoring factor in June 2017 per ML17097A351.


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3.03 ±

Applicability of Laboratory Data to Field-Emplaced Saltstone

NRC expects to close MF 3.03 after NRC determines that representing the hydraulic properties of field-emplaced saltstone with the hydraulic properties of laboratory-produced samples is adequate. That assessment should account for the range of expected disposal conditions of field-emplaced saltstone as well as effects of scale. Alternately, MF 3.03 may be closed if NRC determines that DOE bases the hydraulic properties of saltstone on the properties of an appropriate range of samples of field-emplaced saltstone, rather than on measurements of laboratory-produced samples.

Section 2.3.3.1 - Measure Physical Properties of Laboratory Prepared Saltstone Simulant Samples, Actual Tank 50 Salt Solution Samples, Saltstone In-Line Process Sample, and SDU Cell 2A Emplaced Core Sample Laboratory prepared and process room samples will have physical properties testing performed to determine the hydraulic conductivity, Kd, bulk cured density, porosity, and micro structure/phase analysis. Future testing will compare these properties to those measured from emplaced core sampling.

The saltstone sampling and analysis plan established a strategy for studies to reduce PA uncertainty in the area of SHC, and for correlating grout properties between laboratory-prepared samples and core-drilled samples from actual emplaced grout. A variety of laboratory testing has completed, including one in which sampled were cured under conditions similar to those expected for field-emplaced saltstone. The results were incorporated into the FY2014 SDF SA. Multiple cores were extracted via a wet core drilling process in FY2015 approximately 20 months after the saltstone of interest was processed in the SPF and subsequently emplaced in SDU Cell 2A. The physical property data for SDU-emplaced and laboratory-prepared samples is summarized in the SDU Cell 2A Core Sampling Report (SRR-CWDA-2016-00051), with values for SDF model inputs provided in the report where applicable. Based on these results, NRC narrowed the scope of this monitoring factor in June 2017 per ML17097A351 to the understanding of changes in hydraulic conductivity in the short term between laboratory-prepared and field-emplaced saltstone.


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3.04 ±

Effect of Curing Temperature on Saltstone Hydraulic Properties

NRC expects to close MF 3.04 after NRC determines that projected SDF performance is based on estimates of the hydraulic properties of saltstone (e.g., hydraulic conductivity and diffusivity) that are well-supported. That support should account for the range of curing conditions (i.e., temperatures values, humidity values) experienced by field-emplaced saltstone.

Section 2.3.1 – Critical Property Testing Previous testing and research activities were carried out to define the operating conditions (e.g., water-to-premix ratio, dry feeds variability, and the curing temperature) required to meet or exceed PA expectations of performance.

Saltstone testing of measured hydraulic conductivities is ongoing. A variety of laboratory testing has completed, including one in which samples were cured under conditions similar to those expected for field-emplaced saltstone. The results were incorporated into the FY2014 SDF SA. The physical property data for SDU-emplaced and laboratory-prepared samples is summarized in the SDU Cell 2A Core Sampling Report (SRR-CWDA-2016-00051). Based on the test results, NRC closed this monitoring factor in June 2017 per ML17097A351.

MA 4 – Waste Form Physical Degradation

4.01 ±

Waste Form Matrix Degradation

NRC expects to close MF 4.01 after NRC determines that support for modeled changes in the SHC and diffusivity during the performance period is sufficient.

Section - 2.3.2.2 Studies Related to Cementitious Materials Degradation Due to Radiation Damage A literature search will be conducted to gain a better understanding of the potential degradation of cementitious materials exposed to radiation. and Section 2.2.1 - Prepare Out-year Saltstone Disposal Facility Performance Assessment Revisions This section describes future revisions to the PA that will incorporate improvements to conceptual modeling.

The degradation models for concrete and saltstone grout were revised for the FY2014 SDF SA to incorporate greater conservatisms and to modify inputs to implicitly model fractures in the matrix.

4.02 ±

Waste Form Macroscopic Fracturing

NRC expects to close MF 4.02 after NRC determines that model support for the assumed formation of macroscopic fractures during the performance period is sufficient.

Section 2.3.2 – Degradation Studies Previous testing and research activities were carried out to provide a better understanding of degradation mechanisms and fracturing. [SRNL-STI-2013-00522]

The degradation models for concrete and saltstone grout were revised based on FY2013 test data for the FY2014 SDF SA to incorporate greater conservatisms and to modify inputs to implicitly model fractures in the matrix.


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MA 5 – Waste Form Chemical Degradation

5.01 ±

Radionuclide Release from Field-Emplaced Saltstone

NRC expects to close MF 5.01 after NRC determines that measurements of radionuclide release rates from field-emplaced saltstone used in the PA are reliable.

Section 2.3.2.1 - Long-Term Radiological Lysimeter Program This task is expected to provide Kd values in soil and cementitious materials and additional information about long-term geochemical and transport phenomena that will be used to support the waste release and transport models. Section 2.3.3.1 - Measure Physical Properties of Laboratory Prepared Saltstone Simulant Samples, Actual Tank 50 Salt Solution Samples, Saltstone In-Line Process Sample, and SDU Cell 2A Emplaced Core Sample Laboratory prepared and process room samples will have physical properties testing performed to determine the hydraulic conductivity, Kd, bulk cured density, porosity, and micro structure/phase analysis. Future testing will compare these properties to those measured from emplaced core sampling.

In June of 2015, the NRC issued a TRR titled Technical Review: Oxidation of Reducing Cementitious Waste Forms, Docket No. PROJ0734. The TRR is related to MFs 5.01, 5.02, 5.03, and 5.05. [ML15098A031] Studies to better quantify radionuclide release from field-emplaced saltstone have been complete. Multiple cores were extracted via a wet core drilling process in FY2015 approximately 20 months after the saltstone of interest was processed in the SPF and subsequently emplaced in SDU Cell 2A. The physical property data for SDU-emplaced and laboratory-prepared samples is summarized in the SDU Cell 2A Core Sampling Report (SRR-CWDA-2016-00051), with values for SDF model inputs provided in the report where applicable. Ongoing work related to hydraulic conductivity and Tc-99 and I-129 release from field-emplaced saltstone is discussed in Section 2.3.3.1.


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5.02 ±

Chemical Reduction of Technetium by Saltstone

NRC expects to close MF 5.02 after NRC determines that: (1) model support for the chemical reduction of Tc(VII) to Tc(IV) is robust; and (2) this reduced state is maintained under field conditions. NRC expects that DOE will inform NRC what the ranges of those conditions are expected to be during the performance period.

Section 2.3.1.2 – Contaminant Leaching Characteristics from Saltstone Monolith Studies to support modeled assumptions for Tc release behavior. Section 2.3.1.3 - Determination of Oxidation Front in Saltstone Measurement of cured saltstone grout samples are expected to validate PA assumptions concerning the movement of oxidation fronts through cementitious materials.

In November of 2013, the NRC issued a TRR titled Technical Review: Solubility of Technetium Dioxides in Reducing Cementitious Material Leachates, A Thermodynamic Calculation, Docket No. PROJ0734. The TRR is related to MFs 5.02 and 5.05. [ML13304B159] In June of 2015, the NRC issued a TRR titled Technical Review: Oxidation of Reducing Cementitious Waste Forms, Docket No. PROJ0734. The TRR is related to MFs 5.01, 5.02, 5.03, and 5.05. [ML15098A031] A robust suite of tests have been performed and additional tests are planned in order to develop a detailed understanding of Tc behavior with respect to releases via chemical reductions. Based on the most current data available, the FY2014 SDF SA applied a modified approach for modeling Tc release. Multiple cores were extracted after the saltstone of interest was processed in the SPF and subsequently emplaced in SDU Cell 2A. The physical property data for samples is summarized in the SDU Cell 2A Core Sampling Report (SRR-CWDA-2016-00051) Dynamic leaching and EPA Method 1315 testing were performed in FY2016 and encompassed evaluation of radionuclide-spiked saltstone simulants and actual saltstone cores extracted from SDU Cell 2A. The data from these studies is provided in SREL Doc. R-16-0003. Work continued in FY2017 and is documented in SREL Doc. R-17-0005.


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5.03 ‡

Reducing Capacity of Saltstone

NRC expects to close MF 5.03 after NRC determines that information for the initial reducing capacity of saltstone and the expected evolution of redox conditions over time is adequate.

Section 2.3.1.3 - Determination of Oxidation Front in Saltstone Measurement of cured saltstone grout samples are expected to validate PA assumptions concerning the movement of oxidation fronts through cementitious materials.

In June of 2015, the NRC issued a TRR titled Technical Review: Oxidation of Reducing Cementitious Waste Forms, Docket No. PROJ0734. The TRR is related to MFs 5.01, 5.02, 5.03, and 5.05. [ML15098A031] A review of measured test data resulted in a revised reduction capacity for saltstone (from 0.822 meq e-/g to 0.607 meq e-/g). Pore volume data was revised to apply this more conservative value for the FY2014 SDF SA. In addition, sensitivity models were developed to better understand the effects of varying the initial oxidation conditions in saltstone. These sensitivity models showed that a significant percentage of the saltstone monolith would need to be initially oxidized to significantly alter dose results within 10,000 years.


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5.04 ‡

Certain Risk-Significant Kd Values for Saltstone

NRC expects to close MF 5.04 after NRC determines that model support for the sorption coefficients assumed for radium and selenium for saltstone is adequate. MF 5.04 may be closed based on DOE measurements on either field-emplaced or simulated saltstone. NRC could close MF 5.04 (and open a new MF for selenium) if NRC determines that the inventory of Ra-226 and its ancestors is consistent with the revised inventory assumed in Case K under MFs 1.01 and 1.02.

Section 2.3.1.1 - Measurement of Distribution Coefficients in SRS Subsurface Sediments This study will measure the Kds for various species under oxidizing and reducing conditions in actual subsurface sediments retrieved at SRS (actual SDF soils used). Section 2.3.2.1 - Long-Term Radiological Lysimeter Program This task is expected to provide Kd values in soil and cementitious materials and additional information about long-term geochemical and transport phenomena that will be used to support the waste release and transport models.

In January of 2017, the NRC issued a TRR titled Technical Review: Iodine Sorption Coefficients for Use in Performance Assessments for the Saltstone Disposal Facility, Docket No. PROJ0734. The TRR is related to MFs 5.04, 6.01, 7.01, 10.04, 10.06, and 10.09. [ML16342C575] Due to the relative importance of Kd values, studies are ongoing to improve estimates for site-specific conditions. Studies focus on radionuclides that are expected to contribute significantly to dose risks. The FY2014 SDF SA applied the latest available values and future modeling will consider all available data. In FY2016, DOE produced a comprehensive report of Kd values, including improved documentation for the rationale for values used in modeling. Responses to the NRC RAIs on the FY2014 SDF SA were documented in FY2016, including additional Kd sensitivity analyses. Results indicated that Kd value variability did not impact the ability to meet performance objectives.


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5.05 ‡

Potential for Short-Term Rinse-Release from Saltstone

NRC expects to close MF 5.05 after NRC determines that model support for the exclusion of rinse-release phenomenon from the conceptual model assumed in the DOE 2009 SDF PA is adequate. Alternately, MF 5.05 may be closed after NRC determines that the phenomenon is well-understood and the effect on the projected dose is well supported.

Section 2.3.1.2 – Contaminant Leaching Characteristics from Saltstone Monolith Studies to support modeled assumptions for Tc release behavior.

In November of 2013, the NRC issued a TRR titled Technical Review: Solubility of Technetium Dioxides in Reducing Cementitious Material Leachates, A Thermodynamic Calculation, Docket No. PROJ0734. The TRR is related to MFs 5.02 and 5.05. [ML13304B159] In June of 2015, the NRC issued a TRR titled Technical Review: Oxidation of Reducing Cementitious Waste Forms, Docket No. PROJ0734. The TRR is related to MFs 5.01, 5.02, 5.03, and 5.05. [ML15098A031] A PNNL study found that short-term oxidation of Tc in saltstone can be overcome (i.e., re-reduced) in a fairly short amount of time. In addition, sensitivity models were developed in the FY2014 SDF SA to better understand the effects of varying the initial oxidation conditions in saltstone. Multiple cores were extracted via a wet core drilling process in FY2015 approximately 20 months after the saltstone of interest was processed in the SPF and subsequently emplaced in SDU Cell 2A. The physical property data for SDU-emplaced and laboratory-prepared samples is summarized in the SDU Cell 2A Core Sampling Report (SRR-CWDA-2016-00051), with values for SDF model inputs provided in the report where applicable. Based on these results, NRC narrowed the scope of this monitoring factor in June 2017 per ML17097A351 to the understanding of changes in hydraulic conductivity in the short term between laboratory-prepared and field-emplaced saltstone.


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MA 6 – Disposal Structure Performance

6.01 ‡

Certain Risk-Significant Kd Values in Disposal Structure Concrete

NRC expects to close MF 6.01 after NRC determines that DOE information about radium and selenium sorption in disposal structure concrete is appropriate. Ra-226 information could include either material-specific measurements of radium sorption to disposal structure concrete or additional support for the revised lower inventory estimates for Ra-226 and Th-230 that DOE used in Case K. Se-79 information could include additional model support (e.g., results of laboratory experiments) for the appropriate sorption coefficient for selenium in oxidized disposal structure concrete. Alternately for radium and selenium, if the DOE dose projection changes, NRC could determine that the potential dose from radium and selenium is appropriate without sorption in the disposal structure concrete. DOE may provide additional model support (e.g., results of laboratory experiments) to demonstrate that the sorption coefficient for selenium in oxidized disposal structure concrete reflects the sorption of selenate rather than selenite. For either Ra-226 or Se-79, if appropriate information for one of those radionuclides is provided by DOE, but not the other radionuclide, then NRC could close MF 6.01 and open a new MF for the other radionuclide.

Section 2.3.1.1 - Measurement of Distribution Coefficients in SRS Subsurface Sediments This study will measure the Kds for various species under oxidizing and reducing conditions in actual subsurface sediments retrieved at SRS (actual SDF soils used). Section 2.3.2.1 - Long-Term Radiological Lysimeter Program This task is expected to provide Kd values in soil and cementitious materials and additional information about long-term geochemical and transport phenomena that will be used to support the waste release and transport models.

In January of 2017, the NRC issued a TRR titled Technical Review: Iodine Sorption Coefficients for Use in Performance Assessments for the Saltstone Disposal Facility, Docket No. PROJ0734. The TRR is related to MFs 5.04, 6.01, 7.01, 10.04, 10.06, and 10.09. [ML16342C575] Due to the relative importance of Kd values, studies are ongoing to improve estimates for site-specific conditions. Studies focus on radionuclides that are expected to contribute significantly to dose risks. The FY2014 SDF SA applied the latest available values and future modeling will consider all available data. Responses to the NRC RAIs on the FY2014 SDF SA were documented in FY2016, including additional Kd sensitivity analyses. Results indicated that Kd value variability did not impact the ability to meet performance objectives. The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016 (it was last revised in March 2010). This update incorporated the numerous experiments and geochemical measurements that have been conducted since 2010, resulting in new recommended input values for modeling, integrating recent documented geochemical results, including radionuclide Kd values, solubility values, and cementitious impact factors, and includes a critical evaluation of these values with respect to existing values to assess potential impacts.


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6.02 ±

Technetium Sorption in Disposal Structure Concrete

NRC expects to close MF 6.02 after NRC determines that Kd values for technetium in reduced and oxidized disposal structure concrete are well-supported. Alternately, if the DOE dose projection changes, then NRC could determine that the potential dose from technetium is appropriate without technetium sorption in disposal structure concrete.

Section 2.3.1 - Critical Property Testing A summary of already completed studies support modeled assumptions for Tc sorption.

Based upon new information, the FY2014 SDF SA incorporated a dual dependency model for Tc-99 solute transport (using both the redox state and the solid-phase concentration of Tc-99, via the solubility limit). Additionally, the degradation model for cementitious materials has been revised. Together, these modeling improvements significantly changed the expected transport (and dose results) for Tc-99 within the first 10,000 years after SDF closure.

6.03 ‡

Performance of Disposal Structure Roofs and HDPE/GCL Layers

NRC expects to close MF 6.03 after NRC determines that model support for the amount of water that DOE expects to be diverted by the lower lateral drainage layer, including support for the hydraulic conductivity of the relevant engineered layers, is sufficient. Alternately, NRC could close MF 6.03 if DOE conservatively assumes less diversion around the disposal structures in the PA model.

Section 2.3.2.2 - Studies Related to Cementitious Materials Degradation Due to Radiation Damage A literature search will be conducted to gain a better understanding of the potential degradation of cementitious materials exposed to radiation. and Section 2.3.2.3 - Closure Cap -Term Performance Research will validate assumptions in the PAs concerning the rate of pluggage of the closure cap drainage layer as well as the drainage layer above each SDU.

In April of 2017, the NRC issued a TRR titled Technical Review: Performance of the High Density Polyethylene Layer, High Density Polyethylene/Geosynthetic Clay Liner Composite Layer, and the Lower Lateral Drainage Layer, Docket No. PROJ0734. The TRR is related to MFs 6.03 and 10.02. [ML17081A187] Infiltration into saltstone is influenced by closure cap performance and roof degradation. Closure cap performance is addressed in the discussions associated with MFs 2.01 and 2.02. The roof degradation is addressed through the revised degradation analysis for cementitious materials. The flow cases in the FY2014 SDF SA provide additional insights by varying the infiltration rates and the degradation of cementitious materials. Furthermore, the linear degradation rate for cementitious materials results in less diversion of water around the SDUs. Research was begun by Vanderbilt University and the University of Virginia in FY2017 and will be completed in FY2018 to improve insights related to SDS concrete and HDPE/GCL degradation.


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6.04 ‡

Disposal Structure Concrete Fracturing

NRC expects to close MF 6.04 after NRC determines that support for the amount of fracturing of the disposal structure floor and walls expected to occur during the performance period is adequate or if NRC determines that the estimate that DOE uses in the PA model is conservative.

Sections 2.3.2.4 – Hydraulic Degradation Mechanisms in SDU Concrete Previous testing and research activities were carried out to provide a better understanding of degradation mechanisms and fracturing. [SRNL-STI-2013-00522] Additional research is being carried out to establish key degradation mechanisms for SDU concrete based on long-term, realistic service conditions and incorporating credible, chronic exposure scenarios.

For the FY2014 SDF SA, the model assumes a linear degradation rate for cementitious materials. This assumption conservatively approximates the hydraulic effects of fracturing. The FY2016 SDF SA (SRR-CWDA-2016-00072) addresses cracks in the SDU roof/floor. Research was begun by Vanderbilt University in FY2017 to improve insights in this area. The intent of the work is to gain an enhanced understanding of mechanisms potentially associated with the degradation of SDU concrete and utilize that information to recommend an alternative degradation profile to the currently utilized, and conservative, linear profile. The final report is expected in FY2018.

6.05 ‡

Integrity of Non-cementitious Materials

NRC expects to close MF 6.05 after NRC determines that support for the assumed performance of non-cementitious materials used in the disposal structures is adequate. For example, DOE may perform accelerated testing to estimate long-term performance. Alternately, DOE may be able to use a conservative estimate in the PA model.

N/A

The FY2014 SDF SA modeled non-cementitious materials as gravel and demonstrates that joints have a negligible impact on performance, even when modeled with conservative moisture characteristic curves (MCCs).


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MA 7 – Subsurface Transport

7.01 ‡

Certain Risk-Significant Kd Values in Site Sand and Clay

NRC expects to close MF 7.01 after NRC determines that site-specific measurements for the Kd value for selenium in sand and clay are appropriate. Those measurements should consider the potential effect of the higher pH conditions that are likely to exist downgradient of the SDF. Alternatively, MF 7.01 may be closed if NRC determines that selenium Kd values for SRS sand and clay do not have the potential to significantly affect the dose to an off-site MOP. That determination should consider the uncertainty in other key parameters related to selenium release and transport (i.e., MFs 5.04 and 6.01).

Section 2.3.1.1 - Measurement of Distribution Coefficients in SRS Subsurface Sediments This study will measure the Kds for various species under oxidizing and reducing conditions in actual subsurface sediments retrieved at SRS (actual SDF soils used).

In December of 2016, the NRC issued a TRR titled Technical Review of “Dose Calculation Methodology for Liquid Waste Performance Assessments at the Savannah River Site”, SRR-CWDA-1013-00058, Rev. 1, July 2014, Docket No. PROJ0734. The TRR is related to MFs 7.01, 10.07, 10.08 and 10.09. [ML16277A060] In January of 2017, the NRC issued a TRR titled Technical Review: Iodine Sorption Coefficients for Use in Performance Assessments for the Saltstone Disposal Facility, Docket No. PROJ0734. The TRR is related to MFs 5.04, 6.01, 7.01, 10.04, 10.06, and 10.09. [ML16342C575] Due to the relative importance of Kd values, studies are ongoing to improve estimates for site-specific conditions. Studies focus on radionuclides that are expected to contribute significantly to dose risks. The FY2014 SDF SA applied the latest available values and future modeling will consider all available data. The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016 (it was last revised in March 2010).


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MA 8 – Environmental Monitoring

8.01 § Leak Detection

NRC expects to close MF 8.01 after the leak detection system ends operation or after final waste disposal occurs, whichever comes later.

Section 2.1.3 - Provide General Technical Support on Saltstone Disposal Facility Performance Assessment Issues Activities include supporting NRC on-site observation visits and technical reviews, general project support, testing and research activity support, and development of resolution path forward for NRC open items.

DOE will provide routine/requested information to NRC as it becomes available.

8.02 §

Groundwater Monitoring

NRC does not expect to close MF 8.02 because NRC will monitor groundwater data for the duration of NRC monitoring at the SDF.



MA 9 – Site Stability

9.01 ‡

Settlement Due to Increased Overburden

NRC expects to close MF 9.01 after NRC determines that the projections of settlement in the recent geotechnical investigations will not adversely affect SDF performance. Alternately, DOE may provide NRC information that allows NRC to determine that the new DOE settlement projections are consistent with the values assumed in the DOE 2009 SDF PA.

N/A Geotechnical evaluations for current and planned SDUs indicate that settlement will not be significant.

9.02 ‡

Settlement Due to Dissolution of Calcareous Sediment

NRC expects to close MF 9.02 after NRC assesses a new DOE projection of the likelihood of the formation of sinks during the period of performance at the SDF and any resulting effects on site stability.



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MA 10 – Performance Assessment Model Revisions

10.01 ±

Implement-ation of Conceptual Models

NRC expects to close MF 10.01 after DOE updates the PA and NRC determines that intermediate model results are consistent with the conceptual models, quality assurance methods used are appropriate, and parameter values and uncertainty ranges are appropriate.

Section 2.2.1 - Prepare Out-year Saltstone Disposal Facility Performance Assessment Revisions This section describes future revisions to the PA that will incorporate improvements to conceptual modeling.

The FY2014 SDF SA provides a revised model with a number of important updates. With it, DOE also provides more intermediate results and evidence of quality assurance practices.

10.02 ±

Defensibility of Conceptual Models

NRC expects to close MF 10.02 after DOE updates the PA and NRC determines that the conceptual models are appropriate.

Section 2.2.1 - Prepare Out-year Saltstone Disposal Facility Performance Assessment Revisions This section describes future revisions to the PA that will incorporate improvements to conceptual modeling.

In April of 2017, the NRC issued a TRR titled Technical Review: Performance of the High Density Polyethylene Layer, High Density Polyethylene/Geosynthetic Clay Liner Composite Layer, and the Lower Lateral Drainage Layer, Docket No. PROJ0734. The TRR is related to MFs 6.03 and 10.02. [ML17081A187] The FY2014 SDF SA provides a revised model with a number of important updates. With it, DOE also provides more intermediate results and evidence of quality assurance practices.

10.03 ‡

Diffusivity in Degraded Saltstone

NRC expects to close MF 10.03 after DOE updates the PA and NRC determines that the diffusivity information, including the model of the movement of the oxidation front, is well-supported.

Section 2.3.1.3 - Determination of Oxidation Front in Saltstone Measurement of cured saltstone grout samples are expected to validate PA assumptions concerning the movement of oxidation fronts through cementitious materials.

A review of measured test data resulted in a revised diffusivity for saltstone (from 1.0E-07 cm2/sec to 1.0E-08 cm2/sec). In addition, sensitivity models were included in the FY2014 SDF SA to better understand the effects of varying the initial oxidation conditions in saltstone. These sensitivity models showed that a significant percentage of the saltstone monolith would need to be initially oxidized to significantly alter dose results within 10,000 years.


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10.04 †

Kd Values for Saltstone

NRC expects to close MF 10.04 after DOE updates the PA and NRC determines that the Kd values for saltstone for any radionuclides that become risk-significant in the updated PA are well-supported.

Section 2.3.2.1 - Long-Term Radiological Lysimeter Program This task is expected to provide Kd values in soil and cementitious materials and additional information about long-term geochemical and transport phenomena that will be used to support the waste release and transport models.

In January of 2017, the NRC issued a TRR titled Technical Review: Iodine Sorption Coefficients for Use in Performance Assessments for the Saltstone Disposal Facility, Docket No. PROJ0734. The TRR is related to MFs 5.04, 6.01, 7.01, 10.04, 10.06, and 10.09. [ML16342C575] Due to the relative importance of Kd values, studies are ongoing to improve estimates for site-specific conditions. Studies focus on radionuclides that are expected to contribute significantly to dose risks. The FY2014 SDF SA applied the latest available values and future modeling will consider all available data. The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016 (it was last revised in March 2010).

10.05 †

Moisture Characteristic Curves

NRC expects to close MF 10.05 after DOE updates the PA and NRC determines that the MCCs are well-supported. Alternatively, MF 10.05 may be closed if, in an updated PA, DOE assumes the relative permeability is 1, which means that DOE does not use MCCs in the updated PA.

Section 2.3.1 – Critical Property Testing Previous testing and research activities investigated the impact of curing temperature on the moisture retention properties in saltstone. Characteristic curves for high cure temperature samples were compared to those based on saltstone cured at room temperature.

In March of 2017, the NRC issued a TRR titled Technical Review: Saltstone Waste Form Hydraulic Performance, Docket No. PROJ0734. The TRR is related to MF 10.05. [ML17018A137] The FY2014 SDF SA applied revised MCCs to incorporate data from the latest studies.


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10.06 †

Kd Values for Disposal Structure Concrete

NRC expects to close MF 10.06 after DOE updates the PA and NRC determines that the Kd values for disposal structure concrete for any radionuclides that become risk-significant in the updated PA are well-supported.

Section 2.3.2.1 - Long-Term Radiological Lysimeter Program This task is expected to provide Kd values in soil and cementitious materials and additional information about long-term geochemical and transport phenomena that will be used to support the waste release and transport models.

In January of 2017, the NRC issued a TRR titled Technical Review: Iodine Sorption Coefficients for Use in Performance Assessments for the Saltstone Disposal Facility, Docket No. PROJ0734. The TRR is related to MFs 5.04, 6.01, 7.01, 10.04, 10.06, and 10.09. [ML16342C575] Due to the relative importance of Kd values, studies are ongoing to improve estimates for site-specific conditions. Studies focus on radionuclides that are expected to contribute significantly to dose risks. The FY2014 SDF SA applied the latest available values and future modeling will consider all available data. The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016 (it was last revised in March 2010).

10.07 †

Calculation of Build-Up in Biosphere Soil

NRC expects to close MF 10.07 after DOE updates the PA and NRC determines that the soil Kd values are well-supported in the soil build-up calculation (i.e., if DOE chose conservative low Kd values in the transport calculation, then the soil Kd values may not be the same Kd values used in the transport calculation).


In December of 2016, the NRC issued a TRR titled Technical Review of “Dose Calculation Methodology for Liquid Waste Performance Assessments at the Savannah River Site”, SRR-CWDA-1013-00058, Rev. 1, July 2014, Docket No. PROJ0734. The TRR is related to MFs 7.01, 10.07, 10.08, and 10.09. [ML16277A060] For the FY2014 SDF SA, DOE developed a revised dose calculator incorporating the latest available data related to exposure factors. The revised dose calculator also included a soil-build up calculation that applied the most recent Kd values.


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10.08 ‡

Consumption Factors and Uncertainty Distributions for Transfer Factors

NRC expects to close MF 10.08 after DOE updates the PA and NRC determines that the values of consumption factors and uncertainty distributions for transfer factors are well-supported.

Section 2.2.1 - Prepare Out-year Saltstone Disposal Facility Performance Assessment Revisions This section describes future revisions to the PA that will incorporate the latest available information with respect to consumption factors, transfer factors, and uncertainty distributions.

In December of 2016, the NRC issued a TRR titled Technical Review of “Dose Calculation Methodology for Liquid Waste Performance Assessments at the Savannah River Site”, SRR-CWDA-1013-00058, Rev. 1, July 2014, Docket No. PROJ0734. The TRR is related to MFs 7.01, 10.07, 10.08, and 10.09. [ML16277A060] For the FY2014 SDF SA, DOE developed a revised dose calculator incorporating the latest available data related to exposure factors. The revised dose calculator included improved estimates for consumption factors and uncertainty distributions, developed using recent applicable data.


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10.09 †

Kd Values for SRS Soil

NRC expects to close MF 10.09 after DOE updates the PA and NRC determines that the site-specific Kd values for any radionuclides that become risk-significant in the updated PA are well-supported.


In December of 2016, the NRC issued a TRR titled Technical Review of “Dose Calculation Methodology for Liquid Waste Performance Assessments at the Savannah River Site”, SRR-CWDA-1013-00058, Rev. 1, July 2014, Docket No. PROJ0734. The TRR is related to MFs 7.01, 10.07, 10.08, and 10.09. [ML16277A060] In January of 2017, the NRC issued a TRR titled Technical Review: Iodine Sorption Coefficients for Use in Performance Assessments for the Saltstone Disposal Facility, Docket No. PROJ0734. The TRR is related to MFs 5.04, 6.01, 7.01, 10.04, 10.06, and 10.09. [ML16342C575] Due to the relative importance of Kd values, studies are ongoing to improve estimates for site-specific conditions. Studies focus on radionuclides that are expected to contribute significantly to dose risks. The FY2014 SDF SA applied the latest available values and future modeling will consider all available data. The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016 (it was last revised in March 2010).

10.10 ‡

Far-Field Model Calibration

NRC expects to close MF 10.10 after DOE updates the PA and NRC determines that the far-field model calibration, particularly in the area near the SDF, is adequate.

N/A

No new activities have been performed to address this monitoring factor; however, a recent study of HTF data indicate that additional studies would be unlikely to result in significant changes to the model results.

10.11 ‡

Far-Field Model Source Loading Approach

NRC expects to close MF 10.11 after DOE updates the PA and NRC determines that the far-field source loading approach in the model is adequate.

N/A No new activities have been performed to address this monitoring factor.


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10.12 ‡

Far-Field Model Dispersion

NRC expects to close MF 10.12 after DOE updates the PA and NRC determines that the grid refinement used in any hydrological model supporting the updated PA does not increase modeled dispersion beyond the expected physical dispersion.

N/A

In response to the RAIs for the FY2013 SDF SA, DOE has performed sensitivity modeling to evaluate the potential impacts from dispersivity variability.

10.13 †

Impact of Calcareous Zones on Contaminant Flow and Transport

NRC expects to close MF 10.13 after DOE investigates potential preferential pathways due to subsurface calcareous zones and NRC determines that the DOE representation of any preferential pathways due to calcareous zones is adequate.


MA 11 – Radiation Protection Program

11.01 §

Dose to Individuals During Operations

NRC expects to close MF 11.01 at the end of the institutional control period.



11.02 § Air Monitoring NRC expects to close MF 11.02 at the end of

the institutional control period.



Y = Yes N= No N/A = Not applicable 1 Monitoring Factors are color-coded based on NRC-determined prioritizations in NRC Monitoring Plan. [ML13100A113]. Symbols are included for clarity.

± Red = High † Green = Low ‡ Yellow = Medium § Blue = Periodic

2 NRC expectations are from the NRC SDF Monitoring Plan. [ML13100A113]. Subsequent to issuance of the NRC Monitoring Plan, NRC has issued various TRRs related to SDF monitoring activities. [ML13304B159, ML15098A031, ML16196A179, ML16277A060, ML16342C575, ML17018A137, ML17081A187]


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APPENDIX C

U.S. Nuclear Regulatory Commission Monitoring Items for F-Area Tank Farm and H-Area Tank Farm


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Appendix C: U.S. Nuclear Regulatory Commission Monitoring Items for the F-Area Tank Farm and H-Area Tank Farm Factors are colored by NRC priority and a symbol is included with each MF number to ensure clarity. A legend containing a description of the NRC ranking is provided at the end of the Table.

MF1 Factor NRC Monitoring Activities2 / Recommendation(s) Related PA Maintenance Activities as Described in this Document

MA 1 – Inventory

1.1 †

Final Inventory and Risk Estimates

General NRC Monitoring Activities [ML15238A761] During the monitoring period, NRC staff will review special analyses typically performed at the time of closure of each tank that provide updated inventories and risk estimates for the entire tank farm that is the subject of the special analysis. NRC staff will assess the degree to which DOE demonstrates the tank farm meets the performance objectives with the new projected radionuclide inventories and will assess other PA updates. As part of the evaluation, NRC staff will assess the degree to which DOE’s special analyses evaluate uncertainty in the revised inventory. NRC staff should independently verify whether the change in inventory, or changes to other modeling parameters, are expected to lead to an exceedance of the dose-based performance objectives (i.e., a 0.25 mSv/yr [25 mrem/yr] limit to a MOP under 10 CFR 61.41 or an applied 5 mSv/yr [500 mrem/yr] limit to an intruder under 10 CFR 61.42). NRC staff will review special analyses to ensure intruder risks reported in the tank farm PAs are appropriately assessed and evaluated. This factor can be closed following NRC review of the last tank or equipment-specific special analysis prepared by DOE for FTF and HTF.

Section 3.2.4 – Tank Farm SA Each tank is sampled following waste retrieval operations. For each waste tank an SA will be performed to evaluate the impact of the final residual inventory on the conclusions of the PAs. SAs will be available to the NRC in support of NRC’s monitoring role. An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.

Recommendation [FTF and HTF TERs – As Tanks Are Cleaned and Sampled] In the FTF and HTR TERs, NRC staff recommended DOE sample each tank following waste retrieval operations for the purpose of developing a final inventory. (Duplicate, also applies to MF 1.2)

Section – N/A DOE has committed to sampling of each tank following waste retrieval operations. This activity is covered as part of the individual tank project work scope.

Recommendation [FTF and HTF TERs – When Developing Highly Radioactive Radionuclide (HRR) List and When Characterizing Residuals] In the FTF and HTR TERs, NRC staff recommended that DOE continue to evaluate its HRR list and provide sufficient justification for any changes as additional information becomes available. The HRR list should be evaluated especially where it is used to inform decisions, such as the selection of radionuclides characterized in residual waste, selection of treatment technologies, and the screening of radionuclides for the purpose of detailed PA calculations. (Duplicate, also applies to MF 1.2)

Section – N/A Each tank will be sampled following waste retrieval operations and an SA will be performed to evaluate the impact of the final residual inventory on the conclusions of the applicable PA. Development of the final tank residual inventory, including selection of radionuclides to be analyzed, is covered as part of the individual tank project work scope. This recommendation will be considered in development of future inventory determinations and documentation.


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1.1 †

Final Inventory and Risk Estimates

Recommendation [HTF TER] In the HTF TER, NRC staff recommended DOE revise its annulus inventory assumptions in the HTF PA if plans to clean the annuli of Tanks 9H, 10H, and 14H change.

Section 3.2.4 – Tank Farm SA Each tank is sampled following waste retrieval operations. An SA will be performed to evaluate the impact of the final residual inventory on the conclusions of the applicable PA. This recommendation will be considered in development of future SAs. An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.

Recommendation [ML13273A299] DOE should evaluate whether it has appropriately managed inventory uncertainty.


Recommendation [ML13273A299] DOE should provide a stronger technical basis for projected inventory multipliers.



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1.2 †

Residual Waste Sampling

General NRC Monitoring Activities [ML15238A761] NRC staff will review sampling and analysis plans developed for each tank. NRC’s technical review should include, but may not be limited to, the following considerations: • Consideration of intratank waste variability that is important to the sampling design,

including the basis for assumptions regarding homogeneity and the number of samples to be collected

• Use of floor concentration samples for assigning residual waste inventory for tank walls • DOE’s support for assumptions regarding normality of radionuclide concentration

when developing deterministic and probabilistic inventory parameters • Sampling of HRRs or basis for removal of HRRs from the list of radionuclides to be

sampled In addition to review of sampling and analysis plans, NRC staff also will conduct its own independent assessment to verify the list of HRRs in DOE’s assessment is complete. If additional HRRs are identified, NRC staff will meet with DOE to resolve the discrepancies in the list and suggest actions, as appropriate, that DOE could take to ensure that risks are appropriately assessed and managed. NRC staff will review sampling and analysis plans to ensure all HRRs are sampled or a basis for exclusion of an HRR is provided. This MF can be closed when NRC concludes that DOE has provided sufficient information to support its list of HRRs and following review of the last sampling and analysis plan for a tank and following the last planned onsite observation of sampling of a tank (may occur prior to the last tank or ancillary equipment being sampled).

Section 3.1.3 – Provide General Technical Support on Tank Farm PA Issues Ongoing support of NRC’s monitoring role will be provided as necessary.

Recommendation [FTF and HTF TERs – As Tanks Are Cleaned and Sampled] In the FTF and HTF TERs, NRC staff recommended DOE sample each tank following waste retrieval operations for the purpose of developing a final inventory. (Duplicate, also applies to MF 1.1)

Section – N/A DOE has committed to sampling of each tank following waste retrieval operations. This activity is covered as part of the individual tank project work scope.

Recommendation [FTF TER – As Tanks Are Sampled and SAs are Prepared] In the FTF TER, NRC staff recommended DOE better explain intratank waste variability that influences waste characterization and uncertainty evaluation. NRC’s comments were expressed in the context of Tank 18 sampling, but also pertain to future characterization of other tanks. Specifically, NRC commented on (i) lack of explanation regarding differences between past and current sample variability, (ii) potential lack of consideration and explanation of the unexpectedly high tank wall concentrations for Pu-238, and (iii) lack of basis for assumptions regarding normality of sample concentrations and volume estimates when calculating inventory multiplier to be used in the probabilistic analysis.

Section – N/A Each tank will be sampled following waste retrieval operations and an SA will be performed to evaluate the impact of the final residual inventory on the conclusions of the applicable PA. Development of the final tank residual inventory is covered as part of the individual tank project work scope. This recommendation will be considered in development of future inventory determinations and documentation.


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1.2 †

Residual Waste Sampling

Recommendation [FTF and HTF TERs – When Developing HRR List and When Characterizing Residuals] In the TERs for FTF and HTF, NRC staff recommended DOE continue to evaluate its HRR list and provided sufficient justification for any changes as additional information becomes available. The HRR List should be evaluated especially where it is used to inform decisions, such as the selection of radionuclides characterized in residual waste, selection of treatment technologies, and the screening of radionuclides for the purpose of detailed PA calculations. (Duplicate, also applies to MF 1.1)

Section – N/A Each tank will be sampled following waste retrieval operations and an SA will be performed to evaluate the impact of the final residual inventory on the conclusions of the applicable PA. Development of the final tank residual inventory, including selection of radionuclides to be analyzed, is covered as part of the individual tank project work scope. This recommendation will be considered in development of future inventory determinations and documentation.

Recommendation [HTF TER – When Characterizing Residuals] In the HTF TER, the NRC repeated recommendations from the Inventory TRR for Tanks 5F and 6F related to sampling (ML13085A291). These recommendations include: 1. DOE should consider, in its tank sampling design, historical information on tank

waste receipts, and information related to the alteration and redistribution of waste due to cleaning operations that may impact horizontal and vertical waste heterogeneity,

2. DOE should evaluate the option to composite samples within segments (or strata) to preserve information about segment (or strata) variance,

3. DOE should evaluate and present information on the relative contributions of various forms of uncertainty in its estimation of mean tank concentrations,

4. DOE should also consider how it can better assure sample representativeness by improving tank sampling designs, collection tools and instructions.

Section 3.1.3 – Provide General Technical Support on Tank Farm PA Issues These recommendations apply to the overall waste tank residual sampling program and are not tank-specific. These recommendations will be evaluated as funding becomes available.

1.3 †

Residual Waste Volume

General NRC Monitoring Activities [ML15238A761] DOE indicates its intent to improve the method of estimating residual volumes in its LW PA Maintenance Programs (e.g., SRR-CWDA-2012-00022 and SRR-CWDA-2014-00108). NRC staff will monitor DOE’s progress in this area. NRC staff also will attempt to observe DOE’s use of video and photographic records to develop residual waste volumes during an onsite observation. NRC staff will monitor DOE’s visual inspection of internal surfaces to ensure no significant inventory is overlooked (e.g., Pu-238 on the walls of Tank 18F). In the Tank 16H Final Inventory TRR (ML15301A830), NRC reiterated that DOE needs to improve documentation of its volume estimation approach, as well as validate methods used to estimate the residual volumes whether through sampling, measurement or through more qualitative methods (e.g., visual evidence). This factor will be closed once NRC staff concludes DOE has taken steps to improve the process by which it estimates residual volumes or shows that DOE has appropriately managed volume uncertainty. This factor may be reopened if NRC staff identifies issues with DOE’s approach to developing or considering uncertainty in volumes estimates.



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1.3 †

Residual Waste Volume

Recommendation [FTF and HTF TERs – 1 to 5 Years, Next Tank Mapping] In the FTF TER, NRC staff recommended DOE consider improvements to residual material mapping and consideration of uncertainty in volume estimates. In the HTF TER, NRC repeated recommendations related to volume estimations from the Inventory TRR for FTF Tanks 5 and 6. [ML13085A291] NRC expects DOE to address the following technical concerns when estimating residual tank waste volumes in the future: 1. DOE should better understand the accuracy of mapping team height estimates through

additional field validation activities for a range of solid material heights. 2. DOE should clearly communicate how it determines the size of areas to be mapped

and how it manages uncertainty related to height estimates for discretized areas in its deterministic analysis. Likewise, DOE should clarify how it represents uncertainty in the assignment of high and low end heights to these areas (e.g., does it use a height that is clearly below/above the non-uniform surface of the delineated areas).

3. DOE should consider uncertainty in the volume estimates resulting from the transfer of data from photographic and video evidence to hand contoured maps (and then to Excel spreadsheets with a finer discretization).

4. DOE should be more transparent with respect to its approach to (i) mapping annular volumes including use of a crawler to inspect internal surfaces and (ii) estimating residual waste volumes in ventilation ducts. DOE should consider uncertainty in annulus volume estimates.

In lieu of improving the method by which DOE estimates residual waste volume, DOE could manage inventory uncertainty with conservative estimates (i.e., volume estimates that clearly err on the side of higher values).

Section 3.1.3 – Provide General Technical Support on Tank Farm PA Issues Ongoing support of NRC’s monitoring role will be provided as necessary. These recommendations apply to the overall waste tank residual sampling program and are not tank-specific. These recommendations will be evaluated as funding becomes available. Section – N/A DOE will stay up to date with advances in volume measurement applications.

1.4 *

Ancillary Equipment Inventory

General NRC Monitoring Activities [ML15238A761] DOE indicated, in response to NRC comment (SRR-CWDA-2009-00054), its intent to verify PA assumptions regarding transfer line inventories consistent with Section 8.2, “Further Work,” in DOE’s PA (SRS-REG-2007-00002). NRC staff will meet with DOE to discuss DOE’s schedule for characterization of transfer lines to ensure conclusions regarding the relatively low risk estimates for transfer lines are confirmed. Additionally, transfer line inventories are important for the intruder analysis because DOE assumes an intruder can more easily access the residual inventory in a transfer line than in a tank. NRC staff will monitor DOE’s efforts in this area to ensure the assumed transfer line inventories are sufficiently bounding or that increased risk is assessed. This MF can be closed once NRC staff concludes that DOE characterization has confirmed the low risk of ancillary components.

Section 3.1.3 – Provide General Technical Support on Tank Farm PA Issues Ongoing support of NRC’s monitoring role will be provided as necessary. Section 3.2.4 – Tank Farm SA Prior to closure of transfer lines, an SA will be performed to evaluate the impact of the final residual inventory on the conclusions of the PAs. The residual inventories analyzed will be based on the most recent information available at that time.


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1.4 *

Ancillary Equipment Inventory

Recommendation [FTF and HTF TERs – 1 to 5 Years] DOE indicates, in response to NRC comment (SRR-CWDA-2009-00054) its intent to verify PA assumptions regarding transfer line inventories consistent with Section 8.2, “Further Work” in DOE’s PA (SRS-REG-2007-00002).

Section 3.2.4 – Tank Farm SA Prior to closure of transfer lines, an SA will be performed to evaluate the impact of the final residual inventory on the conclusions of the PAs. The residual inventories analyzed will be based on the most recent information available at that time.

1.5 *

Waste Removal (As It Pertains to ALARA)

General NRC Monitoring Activities [ML15238A761] NRC will evaluate removal to the maximum extent practical (MEP) for each cleaned tank to ensure DOE disposal actions are consistent with as low as reasonably achievable (ALARA) criteria. NRC staff will assess DOE compliance with ALARA objectives through review of DOE documentation completed in conjunction with the federal facility agreement closure process. This factor can be closed once all tanks are cleaned and NRC staff has reviewed DOE documentation of removal to the MEP.


Recommendation [FTF and HTF TERs – As Tanks Are Cleaned] In the FTF TER, NRC staff recommended DOE more fully evaluate costs and benefits of additional HRR removal, including (i) consideration of benefits of additional HRR removal over longer performance periods (and considering uncertainty in the timing of peak doses), (ii) justification for assumptions regarding alternative cleaning technology effectiveness, and (iii) comparison of costs and benefits of additional HRR removal to similar DOE activities. In the HTF TER, NRC staff indicated that DOE provide a clear linkage between the Criterion 2 evaluation and the PA results, including consideration of the long-term risks associated with the HTF facility, and indicated that sufficient detail was not provided in the waste determination to ensure consistent format and appropriate content for future cost-benefit analyses.

Section – N/A Development of cost-benefit analyses related to HRR removal are covered as part of the individual tank project work scope. This recommendation will be considered in development of future cost-benefit analyses.

Recommendation [HTF TER] In the HTF TER, NRC staff indicated that it does not have confidence that DOE has adequately evaluated the risk associated with the projected inventory of the Tank 16H annulus. The NRC staff recommended DOE evaluate a waste release scenario due to groundwater in-leakage into and out of the annular region and contacting the high-solubility waste in the annuli of those tanks.

Section 3.2.3 – Prepare Out-Year HTF PA Revisions This recommendation will be considered in the next PA revision. An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.


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1.5 *

Waste Removal (As It Pertains to ALARA)

Recommendation [ML13080A401] In the Cost Benefit Analysis for Tanks 18 and 19, the NRC noted that many additional costs were due to the length of time that had passed between the decision to cease removal activities and the time at which the cost-benefit analysis was performed. DOE does not expect this lapse in time for future cost-benefit analysis for other tanks.


Recommendation [ML13080A401] In the Cost Benefit Analysis for Tanks 18 and 19, the NRC noted issues with the collective dose comparison, which only included 1 person for 50 years, although NRC also noted problems with use of collective dose.


Recommendation [ML13080A401] In the Cost Benefit Analysis for Tanks 18 and 19, the NRC staff questioned DOE’s criteria that additional waste removal be more cost beneficial than other similar DOE activities.


Recommendation [ML13080A401] In the Cost Benefit Analysis for Tanks 18F and 19F, the NRC staff also questioned DOE’s separate consideration of cost and benefit uncertainty in its sensitivity analysis (cumulative impact of uncertainty in the costs and benefits was not considered). Additionally, higher removal rates (e.g., 75 percent) could have been evaluated in sensitivity analysis.



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MA 2 – Waste Release

2.1 §

Solubility-Limiting Phases/Limits and Validation

General NRC Monitoring Activities [ML15238A761] NRC continues to recommend that DOE design and perform waste release experiments using actual tank residual samples as soon as practical. DOE staff should continue to discuss its plans with NRC to ensure experiments are designed to optimize their potential usefulness in supporting the 10 CFR 61.41 compliance demonstrations. This monitoring activity is considered to be the highest priority by NRC staff at this time from both a timing and importance perspective. NRC staff encourages DOE to continue the work that was initiated in FY2014 with respect to obtaining support for the assumed solubility limits of key radionuclides in tank farm waste. Pending results of the waste release experiments, NRC will evaluate the need for additional experiments for other tank waste. This factor can be closed once DOE provides experimental support for the assumed solubilities of key radionuclides relied on for performance. The results of waste release experiments may inform the extent to which additional recommendations would need to be implemented.

Section 3.1.3 – Provide General Technical Support on Tank Farm PA Issues Testing of actual waste (i.e., Tank 18 residuals) was performed in FY2016. The solubilities of Pu, Np, U, and Tc were tested under simulated waste tank chemistry conditions using Tank 18 residual waste samples, with the results documented in Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2016 Report (SRNL-STI-2016-00432). Data was collected via measured concentrations (or solubilities) over a seven week testing period under different chemical conditions (pH and Eh were varied). The chemical conditions reflect a range of states such that the test results can be used to better understand the impact of transitory waste tank chemical conditions on solubility. As discussed in Evaluation of Waste Release Testing Results against the Tank Farm Performance Assessment Waste Release Model (SRR-CWDA-2016-00086), the experimental results indicate there may be some variance from the actual waste solubilities and the Waste Release Model (WRM) assigned solubilities. For example, Np was in all cases more insoluble than assigned in the WRM. The other three elements (Pu, Tc and U) appeared in most instances to be potentially more soluble than was assumed in the WRM. For example, Pu was relatively insoluble when oxidized, but was still more soluble than calculated by the WRM. Even if the experimental results were conservatively accepted into the WRM, the newly assigned solubilities would have a negligible impact on peak doses in 1,000 or 10,000 years. In FY2017 activities were initiated to perform waste release testing on Tank 12 residuals. Actual testing will take place in FY2018.


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2.1 §

Solubility-Limiting Phases/Limits and Validation

Recommendation [FTF and HTF TERs – Short-to-immediate term] In the FTF TER, NRC staff recommended DOE perform experiments to verify validity of Geochemist’s Workbench calculations used to determine solubility-limiting phases, solubility limits, and chemical transition times. These experiments should study (i) pH and Eh evolution of the grout pore water over time, (ii) controlling solubility-limiting phases, and (iii) static and dynamic leach tests to study the mobility of HRRs, including consideration of alteration of tank residuals following chemical cleaning with reagents, such as oxalic acid. In the HTF TER, NRC staff reiterates its FTF recommendation that DOE conduct waste release experiments to (i) distinguish between releases from high solubility compounds and low solubility compounds via semi-dynamic leach tests and (ii) determine constant concentrations of elements of concern under conditions of exposure to local groundwater and grout leachate via static tests. (Duplicate, also applies to MF 2.2)

Section 3.3.1.1 – Waste Release Studies This task focuses on reducing uncertainty surrounding waste release model assumptions. The first step is to attempt to determine solubility values based on actual waste samples for key radionuclides. Testing of actual waste (i.e., Tank 18 residuals) was performed in FY2016. The solubilities of Pu, Np, U, and Tc were tested under simulated waste tank chemistry conditions using Tank 18 residual waste samples, with the results documented in Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2016 Report (SRNL-STI-2016-00432) and in Evaluation of Waste Release Testing Results against the Tank Farm Performance Assessment Waste Release Model (SRR-CWDA-2016-00086). In FY2017 activities were initiated to perform waste release testing on Tank 12 residuals. Actual testing will take place in FY2018.

Recommendation [ML12272A082] DOE should provide additional information to support assumptions regarding longevity of reducing conditions in the contaminated zone. Recent studies (Cantrell and Williams, 2012) suggest that the reducing capacity of the tank grout could be depleted much earlier than assumed in the FTF PA (SRS-REG-2007-00002) and in more recent plutonium solubility modeling performed for Tank 18 (SRNL-STI-2012-00087). Uncertainty in the normative mineralogy assumed in geochemical modeling should be considered under this action. (Duplicate, also applies to MF 2.2)

Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area pending the final resolution of experimental work performed in support MF 2.1. An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.

Recommendation [ML12272A082] DOE should provide additional support for the assumption that the Eh of infiltrating water will remain below a critical threshold at which plutonium solubility will increase to a risk significant value (e.g., updated geochemical modeling indicates a dramatic increase in plutonium solubility occurs at Eh greater than +0.45 V). Uncertainty in the critical threshold and the Eh of infiltrating groundwater should be considered under this action. (Duplicate, also applies to MF 2.2)

Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area pending final resolution of experimental work performed in support MF 2.1.


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2.2 ‡

Chemical Transition Times and Validation

General NRC Monitoring Activities [ML15238A761] NRC staff will evaluate the efficacy of DOE’s use of two chemical transitions, three chemical states, and no more than three solubilities for each key radionuclide with solubility changes assumed in DOE’s PAs to occur at the same time for each key radionuclide for a given tank type. NRC will perform this evaluation through NRC review of literature, DOE-generated geochemical modeling, or through independent geochemical modeling. NRC staff also may observe DOE experiments related to this MF in conjunction with an onsite observation of the FTF or HTF. This factor can be closed when (i) DOE shows that chemical transition times are no longer important to its compliance demonstration (i.e., predicted dose is less than the dose standards for all time) or (ii) DOE provides adequate experimental support for its assumptions regarding chemical transition times.


Recommendation [FTF and HTF TERs – Short-to-Immediate Term] In the FTF TER, NRC staff recommended DOE perform experiments to verify validity of Geochemist’s Workbench calculations used to determine solubility limiting phases, solubility limits, and chemical transition times. These experiments should study (i) pH and Eh evolution of the grout pore water over time, (ii) controlling solubility limiting phases, and (iii) static and dynamic leach tests to study the mobility of HRRs, including consideration of alteration of tank residuals following chemical cleaning with reagents, such as oxalic acid. In the HTF TER, NRC staff reiterated its FTF recommendation that DOE conduct waste release experiments to (i) distinguish between releases from high solubility compounds and low solubility compounds via semi-dynamic leach tests and (ii) determine constant concentrations of elements of concern under conditions of exposure to local groundwater and grout leachate via static tests. (Duplicate, also applies to MF 2.1)

Section 3.3.1.1 – Waste Release Studies This task focuses on reducing uncertainty surrounding waste release model assumptions. The first step is to attempt to determine solubility values based on actual waste samples for key radionuclides. Testing of actual waste (i.e., Tank 18 residuals) was performed in FY2016. The solubilities of Pu, Np, U, and Tc were tested under simulated waste tank chemistry conditions using Tank 18 residual waste samples, with the results documented in Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2016 Report (SRNL-STI-2016-00432) and in Evaluation of Waste Release Testing Results against the Tank Farm Performance Assessment Waste Release Model (SRR-CWDA-2016-00086). In FY2017 activities were initiated to perform waste release testing on Tank 12 residuals. Actual testing will take place in FY2018.

Recommendation [HTF TER – Immediate-Term] The NRC staff recommends DOE include dissolved oxygen concentrations in its modeling that are consistent with measurements of unimpacted groundwater across SRS or collect additional dissolved oxygen measurements within the HTF at locations and elevations that are in closer proximity to the tanks.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions This recommendation will be considered in the next PA revision.


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2.2 ‡

Chemical Transition Times and Validation

Recommendation [ML12272A082] DOE should provide additional information to support assumptions regarding longevity of reducing conditions in the contaminated zone. Recent studies (PNNL-21723) suggest that the reducing capacity of the tank grout could be depleted much earlier than assumed in the FTF PA (SRS-REG-2007-00002) and in more recent plutonium solubility modeling performed for Tank 18 (SRNL-STI-2012-00087). Uncertainty in the normative mineralogy assumed in geochemical modeling should be considered under this action. (Duplicate, also applies to MF 2.1)

Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area pending final resolution of experimental work performed in support MF 2.1. An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.

Recommendation [ML12272A082] DOE should provide additional support for the assumption that the Eh of infiltrating water will remain below a critical threshold at which plutonium solubility will increase to a risk significant value (e.g., updated geochemical modeling indicates a dramatic increase in plutonium solubility occurs at Eh greater than +0.45 V). Uncertainty in the critical threshold and the Eh of infiltrating groundwater should be considered under this action. (Duplicate, also applies to MF 2.1)

Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area pending final resolution of experimental work performed in support MF 2.1.


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MA 3 – Cementitious Material Performance

3.1 ‡

Hydraulic Performance of Concrete Vault and Annulus (As it Relates to Steel Liner Corrosion and Waste Release)

General NRC Monitoring Activities [ML15238A761] NRC staff will review reports, analog studies, and other information used to support DOE’s assumption regarding initial conditions and performance of the concrete vaults to protect the steel liner and limit releases of radioactivity from the annulus. For example, NRC staff will review annual tank inspection reports that provide information regarding trenching, scarifying, and cracking of the concrete vaults, as well as information about groundwater intrusion into the tank vaults. NRC staff will review reports related to previous events that led to potential releases or groundwater in-leakage through joints or cracks in the concrete vaults. Analog studies could include review and evaluation of information obtained from West Valley or other analog sites to better understand the potential for and rates of corrosion of high-level waste tanks/components, as well as mitigative design measures. As part of this MF, NRC staff also will consider the potential for earlier steel liner failure than assumed in DOE’s PA due to corrosion of steel components (e.g., rebar) in the concrete vaults that are close to the vault surface or that may be physically separated but electrically connected. If DOE performs additional modeling or experiments to study the potential for transport of deleterious species into the tank vaults or the separation of iron dissolution and oxygen reduction and subsequent corrosion of steel liners or tanks, NRC staff will review the documentation or provide input on the design and results of the experiments. Experiments to study steel liner corrosion are expected to be relatively difficult to implement with unknown benefit compared to other experimental investigations recommended in NRC’s TERs and discussed in this Monitoring Plan. Therefore, NRC staff does not consider these experiments to be a high priority at this time. Until such time that DOE provides additional support for the estimated lifetimes of the steel liners, NRC staff (i) will assume steel liners will not be effective at mitigating releases for the long time periods DOE relies on the steel liners for performance in the tank farm PAs and (ii) will investigate the support for the performance of other barriers to ensure performance objectives can be met until such support for steel liner performance is provided. NRC staff will perform routine monitoring of DOE’s reliance on cementitious materials to ensure tank farm PA assumptions regarding the ability of the tank vaults to serve as a recognizable and durable barrier to intrusion are valid. In the Cementitious Barriers Partnership Toolbox (CBP) TRR (ML16196A179), the NRC identified QA gaps and limitations in the documentation associated with software development and in the technical basis for the conceptual and mathematical models for the CBP Toolbox. This MF will be reviewed in conjunction with MF 3.4 and can be closed following closure of tanks at FTF and HTF.



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3.1 ‡

Hydraulic Performance of Concrete Vault and Annulus (As it Relates to Steel Liner Corrosion and Waste Release)

Recommendation [FTF and HTF TERs – Long-Term Activity with need contingent on other factors] In the FTF and HTF TERs, NRC staff recommended DOE consider uncertainty in initial conditions and performance lifetime of concrete vaults, because they impact uncertainty in the calculated steel liner failure times.

Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed.

Recommendation [HTF TER – Short- and Immediate-Term] In the HTF TER, NRC staff indicated that DOE should conduct a more comprehensive analysis of contaminant release from the annular regions of Types I and II tanks. Dose projections from the potential release of the radionuclides in the annuli and sand pads are likely to be very sensitive to several key assumptions, which should be well supported. These assumptions include, but are not limited to, (i) the assumed release scenario; (ii) the chemical composition of the infiltrating water; (iii) the volumetric flow rate through grouted tanks, including shrinkage gaps and cracks; and (iv) the solubility of the annulus and sand pad waste. NRC staff also indicated that if the possibility of rise and fall of the water table in the vicinity of the Types I and II tanks cannot be excluded, DOE should evaluate a scenario where water drains from any gaps in the annulus and sand pad regions. (Also applies to MFs 3.2, 3.3, and 3.5)

Section 3.3.2 – To Be Determined Out-Year Testing An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.


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3.2 ‡

Groundwater Conditioning via Reducing Grout

General NRC Monitoring Activities [ML15238A761] NRC staff will monitor DOE experiments to study the potential for groundwater flow through cracks that may form in the tank grout. NRC staff will monitor DOE experiments or perform its own independent experiments to better understand the nature of flow through the tank grout as it impacts the extent to which infiltrating groundwater interacts with and is conditioned by the tank grout. NRC staff also will review information regarding water table rise to evaluate the likelihood of this alternative conceptual model for waste release. Based on the results of the water table rise investigation, an alternative conceptual model may be proposed for a subset of tanks to assess the impact on the compliance demonstration. Specifically, NRC staff will review historical water table elevation data for wells to assess the likelihood of water table rise above the bottom of the tanks. NRC staff also will review design and construction of any DOE mitigation measures used to ensure that the water table remains below the bottom of the tanks. Under contract with the NRC, Center for Nuclear Waste Regulatory Analyses is conducting experiments to study the extent of groundwater conditioning when flow is primarily through preferential pathways or through a cracked grout specimen. The objectives of these experiments are to understand the extent to which infiltrating water chemistry (primarily pH and Eh) is modified by contact with the tank grout and how the tank grout buffering capacities for pH and Eh may change with contact to infiltrating water. Documentation of results of these experiments are expected in CY2015. NRC will also review documentation provided by DOE to support assumptions regarding the extent of groundwater conditioning for as-emplaced tank grout. NRC staff may conduct the technical review activities in conjunction with an onsite observation to observe any laboratory or field experiments in this area. If results of waste release experiments conducted under MF 2.1 show key radionuclides in waste residuals have sufficiently low solubility when in contact with unconditioned SRS groundwaters, MF 3.2 related to the extent of conditioning (and 2.2 related to the longevity of conditioning) will no longer be needed by DOE to support the compliance demonstration and can be closed. If MF 2.1 results indicate unconditioned flow may lead to unacceptably high doses, then this MF will need to be evaluated by NRC and can be closed after DOE (i) shows matrix flow through the grout will dominate waste release or (ii) provides information to support assumptions regarding the level of groundwater conditioning for degraded (cracked) grout.



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3.2 ‡

Groundwater Conditioning via Reducing Grout

Recommendation [FTF and HTF TERs – Intermediate- to Long-Term Activity with need contingent on other factors] In the FTF and HTF TERs, NRC staff recommended DOE obtain greater support for its assumption regarding flow through the tank grout (i.e., fracture versus matrix) flow as it impacts the timing of chemical transition or time to release of HRRs at risk-significant solubility. If found to be risk-significant, DOE should consider the appropriateness of using MCCs for matrix materials to simulate fracture flow. In the HTF TER, NRC staff recommended DOE provide more support for the assumption that the engineered system will not interfere with the ability of the overlying grout to sufficiently condition the infiltrating water for the fully and partially submerged tanks. (Duplicate, also applies to MF 3.3)

Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area pending final resolution of experimental work performed in support MF 2.1. An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.


Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.


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3.3 †

Shrinkage and Cracking of Reducing Grout

General NRC Monitoring Activities [ML15238A761] NRC staff will review grout formulations, calculations, research, test methods, and results to ensure the disposal facility is designed to minimize fast flow path development. NRC staff may conduct technical reviews in conjunction with onsite observations that could include such activities as video inspections of grout pours, observations of grout tests, and inspection of test specimens. The NRC staff will also continue to monitor the importance of alkali-silica reactivity on cementitious material degradation. The NRC staff will continue to evaluate the potential for shrinkage- and cracking-induced preferential flow through the tank grout. MF 3.3 can be closed when DOE demonstrates (i) preferential fast flow into the waste zone of the tanks or through the waste in the annuli for Types I and II tanks at HTF will not occur or (ii) preferential fast flow into the waste zone of the tanks or through the annuli for Types I and II tanks at HTF will not adversely impact performance (e.g., the performance objective can be met under all chemical conditions as discussed in more detail under MA 2, “Waste Release”).

Section 3.1.3 – Provide General Technical Support on Tank Farm PA Issues Ongoing support of NRC’s monitoring role will be provided as necessary. An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.

Recommendation [ML15238A761] • DOE should consider design measures to minimize the occurrence of negative features,

events, or processes that may promote shrinkage or cracking • DOE should consider removal of in-tank equipment that could lead to development of

shrinkage-induced annuli around equipment or corrosion of steel components and associated cracking due to corrosion product expansion.

• DOE should promote the ability of the grout to fill all void spaces to minimize imperfectly bonded grout seams and voids that may form between grout pours.

• DOE should research and evaluate shrinkage compensating agents for use in its grout formulations to minimize shrinkage, shrinkage gap formation, and creation of annuli and void space within grout.

• DOE should ensure temperature gradients are sufficiently low to prevent excessive thermal cracking. Calculations could be conducted to evaluate potential thermal gradients and/or instrumentation could be used to evaluate as-emplaced thermal evolution of grout.

• DOE should ensure the grout is designed to consider the potential for cracking due to differential settlement.

• It may be useful for DOE to research and deploy methods of detecting early crack development in reducing grout used to fill tanks.

Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area when funding is available.


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3.3 †


Recommendation [FTF and HTF TERs – Intermediate- to Long-Term Activity with need contingent on other factors] In the FTF and HTF TERs, NRC staff recommended DOE obtain greater support for its assumption regarding flow through the tank grout (i.e., fracture versus matrix) flow as it impacts the timing of chemical transition or time to release of HRRs at risk-significant solubility. If found to be risk-significant, DOE should consider the appropriateness of using MCCs for matrix materials to simulate fracture flow. In the HTF TER, NRC staff recommended DOE provide more support for the assumption that the engineered system will not interfere with the ability of the overlying grout to sufficiently condition the infiltrating water for the fully and partially submerged tanks. (Duplicate, also applies to MF 3.2)




Recommendation [ML13269A365] In the Tanks 18 and 19 Grout Documentation TRR, the NRC staff stated that DOE has not provided sufficient information to rule out preferential pathways from its reference case. NRC staff expects DOE to provide additional information related to the extent and performance impact of shrinkage. During its review of tank grouting video, NRC staff observed potential segregation of tank grout during placement that could enhance the extent of shrinkage along the periphery of the Type IV tanks. NRC staff also expects DOE to provide additional information on the potential for thermal cracking of the grout monolith for Tanks 18F and 19F.



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3.3 †


Recommendation [ML12212A192, ML13269A365] DOE indicated that it currently does not have plans to conduct shrinkage testing, but may pursue tests in the future. The NRC staff concur with Stefanko and Langton’s (SRNL-STI-2011-00551) recommendations for testing of admixtures and implementation of measures to help mitigate tank grout shrinkage.

Section 3.3.1.2 – Waste Tank Closure Grout Properties Testing Research was performed in FY2017 to improve insights in this area. The intent of the work was to gain an enhanced understanding of tank specific cementitious material properties. Overall results were inconclusive from the shrinkage testing and no further tests are planned in FY2018.

Recommendation [ML12212A192, ML13269A365] The NRC staff believes DOE’s conclusion that the temperature rise was sufficiently low for bulk grouting of Tanks 18F and 19F based on a 1 cubic yard [0.8 m3] bulk scale-up test and will continue to evaluate this technical issue in future onsite observations. A more detailed thermal analysis that considers the specific grout pour sequence and geometry to determine the potential for thermal cracking of the tank grout would improve model support.


Recommendation [ML12212A192, ML13269A365] DOE did not document in the Final Configuration Report (SRR-CWDA-2012-00170) expected causes for the bulk fill grout deviations to mitigate potential concern over the formation of preferential pathways. The Final Configuration Report does provide documentation regarding potential causes of significant deviations in equipment grout fill volumes, however, estimates of remaining void volumes are not provided (e.g., advance design mixer pump [ADMP]). This information could also be used to improve future volume estimates. NRC staff will continue to monitor DOE estimates of void volumes including whether there is additional information that would support a conclusion that the ADMP void spaces were completely filled.


Recommendation [ML14342A784] In the Tanks 5 and 6 Grout Documentation TRR, the NRC staff stated that DOE has not provided sufficient information and testing to exclude from its reference case preferential flow through the tank grout monolith. During its review of tank grouting video, NRC staff observed potential bleedwater segregation of tank grout during placement that could result in inhomogeneity of the monolith, which can affect flow patterns. The NRC staff will continue to evaluate the potential for shrinkage- and cracking-induced preferential flow through the tank grout.



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3.3 †


Recommendation [ML12212A192, ML13269A365] The NRC staff communicated its concern with the potential formation of cracks in the tank grout due to alkali-silica reactivity (ASR). The NRC staff is concerned that DOE’s criterion for acceptance of vendor supplied granite aggregate relies on short-term alkali reactivity tests (ASTM Standard C1260-14), which is unlikely to predict the occurrence of ASR over the very long period of performance for compliance with the performance objective specified at 10 CFR 61.41. Evaluating potential ASR in tank grouts and its potential effect on long-term performance of the engineered barrier system would improve model support for the performance of the grout and understanding of the potential for cracking of the grout.



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3.4 †

Grout Performance

General NRC Monitoring Activities [ML15238A761] NRC will perform technical review activities related to DOE’s testing and development of grout formulations to meet design specifications. Additionally, NRC will monitor DOE’s efforts to deliver a grout mix of sufficient quality to meet performance assumptions in DOE’s PAs from design to as-emplaced conditions in the field. NRC staff will review relevant procedures and documentation related to such items as grout material procurement, production, testing, acceptance and placement in tank farm components. NRC staff will perform technical review activities in conjunction with onsite observations. Onsite observations will include such activities as observations of grout material storage, tests, and acceptance of grout materials; live video streams of grouting operations; review of archived video footage; review of batch tickets for accepted and rejected loads; tour of the command center; and observation of mock-up tests or visual examination of test specimens. NRC will perform routine monitoring of DOE’s use of grout materials to stabilize high-level waste tanks to ensure tank farm PA assumptions regarding the ability of the grouted tank and vaults to serve as a recognizable and durable barrier to intrusion remain valid. The NRC staff will also continue to monitor void volumes in the emplaced grout to the extent information is available and the impact of limestone additions to the grout mix on pH buffering of water contacting the emplaced grout. NRC continues to monitor the potential for segregation of emplaced grout and its impacts on flow through the grout monolith and waste release. In the Tank 16H and Tank 12H Grouting TRR (ML16231A444), the NRC identified information that would enhance DOE’s demonstration of grout performance: • DOE should take reasonable measures to ensure a sufficient number of cement trucks

are in rotation to optimize grout distribution throughout the tank and minimize mounding.

• DOE should address the potential for either a capillary or permeability barrier to form due to the varying hydraulic conductivity of the clean cap and bulk fill grout used in Tank 16H.

• DOE should provide additional information regarding the quantity and performance impact of the presence of standing water in Tank 12H during grouting.

• DOE should evaluate differences in hydraulic conductivity between the Grade 100 and Grade 120 slag used to fill Tank 12H and any resulting performance impact

NRC staff can close this MF after it completes (i) review of DOE-generated grouting documentation and (ii) monitoring of grouting operations. This MF will be reviewed in conjunction with MF 3.1 and can be closed following closure of tanks at FTF and HTF.



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3.4 †

Grout Performance

Recommendation [ML12212A192, ML13269A365] The Tanks 18 and 19 Grout Documentation TRR stated that the NRC staff is concerned that the use of commercially-available Portland cements in Tanks 18 and 19 that differ from the grout mix considered in the FTF PA (SRS-REG-2007-00002) because of substitution of up to 5% by weight limestone would lower the pH buffering capacity of the grout and could affect the timing of release of key radionuclides. Evaluating the effect of limestone substitution in Portland cement on the pH buffering capacity of the grout and the release of key radionuclides improves model support for the modeling of chemical states and transitions of water contacting the residual waste.


3.5 ‡

Vault and Annulus Sorption

General NRC Monitoring Activities [ML15238A761] NRC staff will monitor DOE efforts to study basemat sorption for plutonium and neptunium. NRC staff will review documentation and any analog studies that may provide additional information regarding the ability of the concrete basemats to attenuate release from the tanks or annuli of Types I and II tanks at HTF, including information regarding groundwater inleakage and release from construction joints or other discrete features such as those implicated in the release from HTF Tank 16H. This MF can be closed when (i) sufficient information is available to support assumptions regarding attenuation of key radionuclides (e.g., plutonium, neptunium, cesium, strontium) in the basemats, vaults, or annular grout or (ii) DOE provides sufficient information to show that doses from key radionuclides will be below the dose limits prescribed in the performance objectives with little to no performance from the concrete basemats and vaults and annular grout (e.g., solubility limits for unconditioned groundwater are sufficiently low or natural attenuation of key radionuclides is sufficiently high to compensate for underperformance of the concrete basemat and vaults and annular grout).





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3.5 ‡

Vault and Annulus Sorption

Recommendation [FTF and HTF TERs – Intermediate-Term] In the FTF and HTF TERs, NRC staff commented that given the wide range of values in the literature, NRC staff recommends DOE obtain additional support for basemat Kds for plutonium and neptunium, including consideration of solubility affects from previous evaluations and representativeness of experimentally derived values for aged concrete. DOE should continue to evaluate the appropriateness of selected transport parameters (e.g., cementitious material and soil Kds) and the selection of sorption models during the monitoring period.

Section 3.3.2 – To Be Determined Out-Year Testing The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016 (it was last revised in March 2010). This update incorporated the numerous experiments and geochemical measurements have been conducted since 2010, resulting in new recommended input values for modeling. The revised geochemical data package integrates recent documented geochemical results, including radionuclide Kd values, solubility values, and cementitious impact factors, and includes a critical evaluation of these values with respect to existing values to assess potential impacts. No additional work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area when funding is available.

3.6 *

Waste Stabilization (As it Impacts ALARA)

General NRC Monitoring Activities [ML15238A761] NRC staff will review use of stabilizing materials to determine if DOE has made a reasonable effort to optimize mixing or encapsulating the waste with the stabilizing material. NRC staff will evaluate DOE’s use of stabilizing materials to grout features of the tank and vault system that might otherwise lead to preferential flow through the engineered system and into the environment (e.g., grouting of leak detection channels and sumps contained within the concrete basemats). NRC staff will conduct technical reviews and onsite observations under MFs 3.1 to 3.5, bearing in mind the additional function of the stabilizing grout to maintain doses ALARA. NRC staff will close this factor when MFs 3.1 through 3.5 are closed, and if NRC staff finds DOE’s use of stabilizing cementitious materials consistent with ALARA criteria.


No specific NRC staff recommendations identified. Section – N/A


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MA 4 – Natural System Performance

4.1 §

Natural Attenuation of Key Radionuclides

General NRC Monitoring Activities [ML15238A761] NRC will monitor DOE’s efforts to develop site-specific sorption coefficients that consider the impact of cement-impacted leachate released from the tanks. NRC staff will review any DOE-generated reports or other documentation that provides additional information related to site-specific niobium Kd values. Technical review activities may be conducted in conjunction with onsite observations of any experiments developed to study the attenuation of plutonium, niobium, and other key radionuclides in SRS soils. NRC staff will review information generated by DOE and perform independent modeling to assess whether more mobile forms of plutonium, if evaluated explicitly in DOE’s PA modeling, could reach the inadvertent intruder point of compliance (POC) within 10,000 years. This MF can be closed when NRC staff concludes that DOE has adequately assessed the timing and magnitude of Pu-239 release and transport to the 1-m [3.28 ft] POC and DOE provides support for its treatment of plutonium, niobium, and other key radionuclide sorption in the subsurface at FTF or DOE shows that plutonium, niobium, and other key radionuclide sorption in the subsurface is not needed to support its compliance demonstration (e.g., solubility control effectively limits plutonium releases into the natural environment to non-risk-significant levels).

Section 3.1.3 – Provide General Technical Support on F-Area Tank Farm Performance Assessment Issues Ongoing support of NRC’s monitoring role will be provided as necessary.


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4.1 §


Recommendation [FTF and HTF TERs – Short- and Intermediate-Term] In the FTF TER, NRC staff recommended DOE evaluate appropriateness of averaging Kds of multiple oxidation states to simulate the transport of plutonium in the natural system. Consistent with the recommendation in the FTF TER, in the HTF TER, NRC staff indicated that a more accurate representation of the transport of multivalent plutonium would be to treat the two species separately, assuming the oxidation state distribution could be reasonably quantified. In the HTF TER, NRC staff also questioned the basis for the sandy sediment Kd for Pu of 650 mL/g derived from SRNL-STI-2011-00672, as well as the cement leachate factors that were derived based on Hanford data (SRNL-STI-2009-00473).

Section 2.3.2.2 – Long-Term Radiological Lysimeter Program This task is expected to provide Kd values (including for plutonium) in soil and cementitious materials and additional information about long-term geochemical and transport phenomena that will be used to support the waste release and transport models. The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016 (it was last revised in March 2010). This update incorporated the numerous experiments and geochemical measurements have been conducted since 2010 (including Lysimeter Program testing results), resulting in new recommended input values for modeling. The revised geochemical data package integrates recent documented geochemical results. Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA includes sensitivity analyses that provided additional information regarding the sensitivity of HTF dose results to Pu soil Kd values. The next PA revisions will include more analysis of plutonium attenuation. It is expected that sensitivity analyses will show that peak dose results within the performance period are not sensitive to the Natural Attenuation of plutonium given the SA sensitivity analyses.


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4.1 §


Recommendation [ML13273A299] The Tanks 5F and 6F Special Analysis TRR stated that additional information related to the niobium Kd is needed to have reasonable assurance that DOE disposal actions at the FTF will meet the performance objectives in 10 CFR Part 61, Subpart C.

Section 2.3.1.1 – Measurement of Distribution Coefficients in SRS Subsurface Sediments Testing being performed in support of the SDF PA will be used for the FTF and HTF PAs as applicable. The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016. Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions The next PA revision will include more analysis of niobium attenuation. It is expected that sensitivity analyses will show that peak dose results within the performance period are not sensitive to the Natural Attenuation of niobium.

Recommendation [ML12272A124] The Environmental Monitoring TRR stated that the NRC staff will continue to monitor the Kd averaging approach used to simulate plutonium transport in the natural system at FTF. DOE could address the issue by modeling explicitly more mobile and less mobile forms of plutonium in future performance assessment calculations.

Section 3.3.2 – To Be Determined Out-Year Testing An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA includes sensitivity analyses that provided additional information regarding the sensitivity of HTF dose results to Pu soil Kd values. The next PA revisions will include more analysis of plutonium attenuation. It is expected that sensitivity analyses will show that peak dose results within the performance period are not sensitive to the Natural Attenuation of plutonium given the SA sensitivity analyses.


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4.1 §

Natural Attenuation of

Key Radionuclides

Recommendation [ML12272A124] The Environmental Monitoring TRR stated that the NRC staff will continue to monitor support for cement leachate factors developed for plutonium (and other constituents). DOE could provide support for cement leachate factors by performing site-specific analyses.

Section 3.3.2 – To Be Determined Out-Year Testing The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016. No additional work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area when funding is available.

Recommendation [ML12272A124] The Environmental Monitoring TRR stated that the NRC staff will continue to monitor the basis for selection of the niobium Kd value of 160 L/kg used in the Tanks 5F and 6F Special Analysis. DOE could address the technical issues by verifying the batch experiments did not exceed solubility limits and are representative of conditions at FTF (e.g., plot solid phase versus aqueous phase concentration or Kd versus concentration; evaluate Kd for FTF aquifer soils) or perform additional experiments to verify the niobium Kd.


Recommendation [ML116277A060] As a result of its review of the revised dose methodology for Liquid Waste PAs (SRR-CWDA-2013-00058, Revision 1), NRC identified a number of items that should be addressed in future HTF and FTF PAs revisions. The complete listing of items is documented in the Dose Calculation Methodology TRR.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions These items will be considered in the next PA revision.

4.2 †

Calcareous Zone Characterization

General NRC Monitoring Activities [ML15238A761] NRC staff should observe field tests and review and evaluate results of tracer tests and field mapping DOE may conduct to ascertain the significance of existing calcareous “soft” zones on flow and transport from the tank farms. NRC Staff will review relevant geotechnical logs acquired in the vicinity of tank farms to stay informed of the potential for and characteristics of soft zones that may be identified in the future. Finally, if DOE opts to employ downhole visualization or other methods to monitor local groundwater velocities associated with soft zones, NRC staff will review and evaluate DOE’s analysis of these data. NRC may conduct technical review activities in conjunction with onsite observations of field activities, such as calcareous zone outcrop mapping on Upper Three Runs Creek. This factor will be closed when DOE has provided NRC sufficient information to show its treatment of calcareous zones in the tank farm PAs is reasonable or adequate to assess risk.



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4.2 †

Calcareous Zone Characterization

Recommendation [ML15238A761] DOE could monitor flow velocities at screen levels both consistent and inconsistent with known existing soft zones to assess local fast flow path gradients of soft zones to provide additional confidence that current PA groundwater modeling treatment is acceptable.


Recommendation [FTF and HTF TERs – Next PA Update (Long-Term)] In the FTF and HTF TERs, NRC staff recommended DOE continue to evaluate significance of calcareous zone dissolution on FTF flow and transport, including conduct of tracer studies and field mapping of seepage locations along Upper Three Runs Creek. Site-specific Kds may also need to be developed for the Upper Three Runs Aquifer-Lower Zone. (Duplicate, also applies to MF 6.2)


4.3 †

Environmental Monitoring

General NRC Monitoring Activities [ML15238A761] NRC staff will review any data collected by DOE for the tank farms for the purpose of evaluating disposal facility performance. NRC staff will review and evaluate groundwater monitoring data as a technical review activity under this factor. NRC staff will continue to review the adequacy of the tank farm monitoring well network with respect to its ability to detect releases from the tank farms. NRC may conduct technical review activities for this monitoring activity in conjunction with onsite observations related to groundwater sampling, well construction, and other field activities. SCDHEC oversight may be leveraged in this area to ensure the quality of data collected. MA 4 will be renamed “Environmental Monitoring” once MF 4.1 and 4.2 have been closed. MA 4 will remain open indefinitely.


Recommendation [ML12272A124] The Environmental Monitoring TRR stated that the NRC staff will continue to evaluate the source of elevated Tc-99 levels in well FTF 28. It is not clear that releases from the F-Area Inactive Process Sewer Line could migrate vertically to the lower zone of the Upper Three Runs Aquifer in such a short distance from the source. This evaluation is important to ensure that the hydrogeological system at FTF is well understood and that releases from the tanks could be detected by the monitoring well network. DOE could provide additional support for the source of contamination detected at well FTF 28 by performing particle tracking to better understand contaminant plume trajectories. DOE could also perform a more formal statistical analysis of FTF and Western Groundwater Operable Unit well data to correlate contaminant concentrations associated with various sources.



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4.3 †

Environmental Monitoring

Recommendation [ML12272A124] The Environmental Monitoring TRR stated that the NRC staff will continue to monitor the ability of the tank farm monitoring well network to detect releases from the tank farm facilities following closure. DOE could (i) evaluate the monitoring well network by performing an analysis of the centerline of plumes emanating from tank sources should releases occur in the future and (ii) provide input on optimal well locations to ensure that future releases from the tank farm facility would be detected.


MA 5 – Closure Cap Performance

5.1 *

Long-Term Hydraulic Performance

General NRC Monitoring Activities [ML15238A761] NRC will monitor additional information to support the assumed long-term hydraulic conductivity of the foundation layer. In addition, NRC staff will monitor construction quality and settlement at the tank farms to help ensure assumed performance of the high density polyethylene/geosynthetic clay liner (HDPE/GCL) composite layer is not adversely impacted. NRC will monitor the quality assurance/quality control for closure cap construction and settlement data collected during tank farm operations as well as nearby facilities. NRC also will review relevant studies and tests related to HDPE/GCL performance. This factor can be closed after DOE’s construction of the closure cap and demonstration of its hydraulic performance.


Recommendation [FTF and HTF TERs – Long-Term Activity] In the FTF and HTF TERs, NRC staff recommended DOE provide additional model support for (i) the long-term hydraulic conductivity of the upper foundation layer and lateral drainage layer and (ii) the long-term erosion of the topsoil layer. (Duplicate, also applies to MF 5.2)

Section 3.3.2 – To Be Determined Out-Year Testing Work was begun by the University of Virginia in FY2017 related to degradation of the closure cap layers and resulting infiltration rates based on new field information and calculation methodologies. Work will complete in FY2018.

Recommendation [FTF and HTF TERs – Long-Term Activity] In the FTF and HTF TERs, NRC staff recommended DOE conduct a preliminary evaluation of erosion protection designs (e.g., assessment of an acceptable rock source, and the ability of an integrated drainage system to accommodate design features) prior to completing the final closure cap design. (Duplicate, also applies to MF 5.2)

Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area closer to the time when final closure cap design is being developed.


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5.2 *

Long-Term Erosion Protection Design

General NRC Monitoring Activities [ML15238A761] NRC staff will review and evaluate information pertaining to erosion processes of the vegetative and topsoil layers, including cover maintenance activities. If DOE performs simulations of the influence of clogging and ponding in the perimeter drainage structures on flow in the vadose zone, NRC will review results of these simulations to evaluate risk significance of the uncertainties in the long-term performance of the perimeter drainage structure. NRC will specifically monitor use of the engineered closure cap as a barrier to intrusion. This factor can be closed after DOE’s construction of the closure cap and demonstration of its physical stability.


Recommendation [ML15238A761] NRC staff recommends that DOE provide additional support for the long-term erosion of the topsoil layer and conduct a preliminary evaluation of erosion protection designs.

Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area closer to the time when final closure cap design is being developed. Work was begun by the University of Virginia in FY2017 related to degradation of the closure cap layers and resulting infiltration rates based on new field information and calculation methodologies. Work will complete in FY2018.

Recommendation [FTF and HTF TERs – Long-Term Activity] In the FTF and HTF TERs, NRC staff recommended DOE provide additional model support for (i) the long-term hydraulic conductivity of the upper foundation layer and lateral drainage layer and (ii) the long-term erosion of the topsoil layer. (Duplicate, also applies to MF 5.1)



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5.2 *

Long-Term Erosion Protection Design

Recommendation [FTF and HTF TERs – Long-Term Activity] In the FTF and HTF TERs, NRC staff recommended DOE conduct a preliminary evaluation of erosion protection designs (e.g., assessment of an acceptable rock source, and the ability of an integrated drainage system to accommodate design features) prior to completing the final closure cap design. (Duplicate, also applies to MF 5.1)


5.3

Closure Cap Functions That Maintains Doses ALARA

General NRC Monitoring Activities [ML15238A761] NRC staff will monitor DOE’s disposal actions as they pertain to tank farm closure cap design, construction, and maintenance consistent with ALARA criteria. This monitoring area will remain open throughout DOE’s development, construction, and completion of final closure caps, unless final design information indicates the MFs are not risk significant. When DOE develops final closure cap designs, NRC will revise the Monitoring Plan, as appropriate, to describe the monitoring activities relevant to the final designs. NRC staff will monitor DOE’s development of specific designs for the closure caps and determine whether these designs are likely to significantly alter DOE and NRC conclusions regarding the conceptual design analyzed in the PA. Prior to any construction activities, NRC staff will review specifications for closure cap construction materials and quality assurance/quality control procedures for assuring these materials meet specifications. During construction, NRC staff should observe the placement of these materials and the quality control testing to assure the as-built closure cap will meet design specifications. NRC staff also will evaluate available data from similar covers built on the larger SRS site and other humid sites.

Section 3.1.3 – Provide General Technical Support on Tank Farm PA Issues Ongoing support of NRC’s monitoring role will be provided as necessary. Work was begun by the University of Virginia in FY2017 related to degradation of the closure cap layers and resulting infiltration rates based on new field information and calculation methodologies. Work will complete in FY2018.


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MA 6 – Performance Assessment Maintenance

6.1 #

Scenario Analysis

General NRC Monitoring Activities [ML15238A761] NRC staff will review PA revisions to evaluate adequacy of scenarios considered. Specifically, NRC staff will review the DOE methodology for identification, screening, and dispositioning of Features, Events, and Processes (FEPs) and the formation of scenarios considered in the PAs. NRC staff should verify FEPs identified by DOE, including all FEPs having a potential to influence compliance with performance objectives. NRC will also evaluate DOE’s consideration of FEPs related to inadvertent intrusion. NRC staff should examine the technical basis for screening FEPs from further consideration in the PA. NRC staff also should examine DOE bases for the formation of scenarios considered in the PAs to determine whether they include all FEPs that have not been screened from further consideration. NRC will close this factor when DOE demonstrates that all risk-significant FEPs have been (or will be under another MF) adequately evaluated in PA documentation.


Recommendation [FTF and HTF TERs – Next PA Update] In the FTF TER, NRC staff recommended DOE perform a systematic scenario analysis in which FEPs are identified, screened, and dispositioned using transparent and traceable documentation of the FEPs considered, the screening arguments, and how FEPs are implemented in the models to support future waste determination efforts. DOE performed a FEPs analysis to support the final waste determination for FTF. NRC staff reviewed the FEPs analysis and documented the results of its review in a TRR. (ML13277A063; see also Table B-1) In the HTF TER, similar to the findings in the TRR for FTF, NRC staff recommended that DOE include subject matter experts on the screening team in the specific engineering and scientific disciplines that are pertinent to the professional judgments being made. NRC staff noted that the screening documentation could be more transparent. The NRC staff also recommended that DOE improve the transparency and traceability of its implementation of FEPs to ensure comprehensive, accurate, and traceable links to clear descriptions of how included FEPs are actually implemented in the HTF PA.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions The next PA revision will incorporate the results of FEP analysis updates and any additional information relative to FEPs available at that time.


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6.1 #

Scenario Analysis

Recommendation [ML13277A063] In the Review of FEPs TRR, the NRC staff’s review of the DOE screening methodology finds that DOE properly focused on likelihood and impact as criteria for screening, but identifies several concerns with DOE’s screening of FEPs, including the membership of the FEPs screening team and the documentation of each subject matter expert’s basis for judgment. The NRC staff’s review also identifies questions with the screening process for selected FEPs. Finally, the NRC staff’s review finds that DOE’s crosswalk of included FEPs has the potential to enhance transparency and traceability, while NRC staff identifies multiple examples where transparency and traceability are reduced, which results in a loss of confidence that all relevant FEPs are adequately considered in the FTF Performance Assessment.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions The next PA revision will incorporate the results of FEP analysis updates and any additional information relative to FEPs available at that time.

6.2 #

Model and Parameter Support

General NRC Monitoring Activities [ML15238A761] NRC staff will review DOE’s PA revisions to evaluate the selection of models and justification of parameters. Specifically, NRC staff will examine information DOE generates, including experimental and site characterization data and information from literature, to support model selection and justify parameters. NRC staff also will review DOE methods to characterize data and model uncertainty and propagate the uncertainty through the PAs. NRC staff will use a graded approach to focus on aspects of most importance to demonstrating compliance with the performance objectives. This factor can be closed when DOE provides sufficient information to support risk-significant models and/or model parameters listed in Appendix A of the NRC Monitoring Plan related to MF 6.2.


Recommendation [FTF and HTF TERs – Next PA Update (Long-Term)] In the FTF and HTF TERs, NRC staff recommended DOE continue to evaluate significance of calcareous zone dissolution on FTF flow and transport, including conduct of tracer studies and field mapping of seepage locations along Upper Three Runs Creek. Site-specific Kds may also need to be developed for the Upper Three Runs Aquifer-Lower Zone. (Duplicate, also applies to MF 4.2)


Recommendation [FTF TER – Next PA Update] As documented in the FTF TER, DOE will explain the differences in the inventory lists for tanks versus ancillary equipment in future PA documentation. DOE made this commitment in an NRC/DOE teleconference on the FTF RAIs held on June 28, 2011.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions The next PA revision will explain the differences in the inventory lists for tanks versus ancillary equipment.


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6.2 #


Recommendation [HTF TER – Next PA Update] In the HTF TER, NRC staff repeated an FTF TRR (ML13273A299) comment indicating that DOE should provide a stronger technical basis for the projected inventory multipliers used in the probabilistic analysis. Given the significant fraction of the Tank 5F and 6F radionuclide inventories that were underestimated, it was not clear to the NRC staff that the inventory multipliers should be biased at 100 times less and only 10 times higher. NRC staff went on to state that DOE should analyze trends in projections versus actual inventories by radionuclide to update the multiplier assumptions for the probabilistic analysis.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions The next PA revision will explain the differences in the inventory lists for tanks versus ancillary equipment.

Recommendation [FTF and HTF TERs – Long-Term Activity with need contingent on other factors] In the FTF TER, NRC staff recommended DOE consider uncertainty in steel liner performance, including more aggressive service conditions and corrosion mechanisms than assumed in the PA, as well as a patch model for waste release, if deemed to be risk significant. In the HTF TER, similar to previous FTF consultative comments, NRC staff questioned DOE’s assumed time-invarient oxygen diffusivity of 10−6 cm2/s given expected degradation of concrete vaults over time and potential presence of bypassing pathways through the system.


Recommendation [FTF and HTF TERs – Intermediate-Term] In the FTF TER, NRC staff recommended DOE obtain additional support for probabilistic parameter distributions, including solubility limiting phases, cement Kds (based on sediment variability), chemical transition times, basemat bypass, and configuration probability. In the HTF TER, NRC staff recommended DOE incorporate in probability distributions “pessimistic” values that exceed base case solubility limits and that DOE obtain support for the solubility and probability assignments


Recommendation [FTF TER – Long-Term] In FTF TER, NRC staff recommended DOE acquire FTF specific data to support material property assignments, including hydraulic conductivity, MCCs, and KdS.


Recommendation [HTF TER – Prior to final closure] In the HTF TER, NRC staff recommended DOE provide additional model support to understand the effects of perimeter infiltration and focused infiltration in the drainage valley between the East and West Caps on near-field and far-field groundwater flow patterns and radionuclide transport. The analysis should include appropriate refinement of the grid cells receiving recharge and a well-supported value for the diversion of flow.



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6.2 #


Recommendation [FTF and HTF TERs – Long-Term] In the FTF TER, NRC staff indicated that it would monitor DOE’s efforts to study the impact of cement leachate on radionuclide mobility. NRC reviewed cement leachate factors utilized in the HTF PA and listed several technical concerns in the HTF TER, most notably the lack of site-specific information and basis for some of the factors. In the Tank 16H Special Analysis TRR (ML15301A710), the NRC indicated that it needs DOE to provide additional information to support its selection of the iodine distribution coefficient.

Section 2.3.1.1 – Measurement of Distribution Coefficients in SRS Subsurface Sediments The geochemical data package (SRNL-STI-2009-00473) used to identify a justified set of geochemical data inputs for the various transport modeling at SRS was updated in July 2016 (it was last revised in March 2010). This update incorporated the numerous experiments and geochemical measurements have been conducted since 2010, resulting in new recommended input values for modeling. The revised geochemical data package integrates recent documented geochemical results, including radionuclide Kd values, solubility values, and cementitious impact factors, and includes a critical evaluation of these values with respect to existing values to assess potential impacts. Distribution Coefficient studies measured the Kds for various species under oxidizing and reducing conditions in actual subsurface sediments retrieved at SRS (actual SDF soils used). This will be applicable to FTF and HTF modeling as well. Results are discussed in Section 2.3.1.1. Section 3.3.2 – To Be Determined Out-Year Testing No work related to this recommendation is currently being performed. DOE will evaluate potential activities in this area when funding is available.

Recommendation [FTF TER – Next PA Update] In the FTF TER, NRC staff recommended DOE address the significant amount of dispersion evident in its near-field and far-field PORFLOW models, including evaluation of the need for mesh refinement to ensure that contaminant plumes are not artificially dispersed over the volume of the cells in the far-field model. Nonphysical dispersion may be attributable to large changes in adjacent element size and large differences in element sizes between the vadose zone and far-field models. DOE should evaluate the adequacy of the time discretization of the model(s) for swiftly moving constituents such as Tc-99.



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6.2 #


Recommendation [FTF TER – Next PA Update] In the FTF TER, NRC staff recommended DOE evaluate the appropriateness of the assumed level of physical dispersion in the FTF model (i.e., dispersivities).


Recommendation [FTF and HTF TERs – Next PA Update] In the FTF TER, NRC staff recommended DOE provide greater transparency and traceability of far-field model calibration, including consideration of more extensive calibration focused strictly on the area of interest. In the HTF TER, NRC staff made recommendations similar to those in the FTF TER, but more strongly indicated that the model may not be sufficiently calibrated local to HTF, and recommended specifically that DOE study uncertainty in calibration targets and provide support for hydraulic conductivity assignments (Kh was artificially lowered in elliptical regions during the calibration process), including consideration of conducting pumping tests to provide support for the model and model parameters.


Recommendation [FTF and HTF TERs – Next PA Update] NRC staff indicated in the FTF TER that Gordon Aquifer concentrations should not be used to demonstrate compliance with the performance objectives if higher concentrations are observed in another aquifer that can support groundwater-dependent pathways. These statements were repeated in the HTF TER.


Recommendation [HTF TER– Next PA Update] In the HTF TER, NRC staff recommended DOE evaluate the compliance boundary and loading of the contaminant source cells (i.e., tank cells in the far-field model) to ensure that the dose estimates are not significantly underestimated.


Recommendation [FTF and HTF TERs – Next PA Update] In the FTF TER, NRC staff recommended DOE evaluate plant transfer factor uncertainty in future updates to its PA. DOE should consider the appropriateness of excluding common vegetable types in its assignment of plant transfer factors (DOE only considers root vegetable data) based on production data rather than household data that might be more appropriate for a resident gardener. In the HTF TER, NRC staff indicated that DOE addressed the use of root vegetable transfer factors; however, uncertainty in plant transfer factors was not addressed.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions A review of the current HTF Base Case conceptual model inputs was performed in 2016 to identify inputs requiring revision (SRR-CWDA-2015-00158), including a review of the dose calculator inputs (SRR-CWDA-2013-00058). This recommendation will be considered further in the next PA revision.


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6.2 #


Recommendation [FTF and HTF TERs – Next PA Update] In the FTF TER, NRC staff recommended DOE evaluate appropriateness of assumptions related to drinking water consumption in future updates to its PA, such as partitioning consumption rates based on use of both bottled and community water. Biosphere parameters should be reasonably conservative and reflect the behavior of the average member of the critical group. NRC staff reiterated the FTF TER recommendation in the HTF TER.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions A review of the current HTF Base Case conceptual model inputs was performed in 2016 to identify inputs requiring revision (SRR-CWDA-2015-00158), including a review of the dose calculator inputs (SRR-CWDA-2013-00058). This recommendation will be considered further in the next PA revision.

Recommendation [FTF TER – Next PA Update] In the FTF TER, NRC staff indicated that DOE better assess uncertainty in the timing of peak dose, given the inherent level of uncertainty associated with predicting doses over tens of thousands of years. Additionally, NRC staff indicated that key parameters, such as steel liner failure times and chemical transition times, may be overly constrained.


Recommendation [FTF and HTF TERs – Next PA Update] In the FTF TER, NRC staff recommended DOE provide additional support for the likelihood of its base case or expected Case A. In the HTF TER, NRC staff went on to state the NRC staff thinks that additional information is needed to support the compliance case, Case A. Ideally, supporting information would be in the form of additional experimental or field data, natural analogs, peer review, expert elicitation, and other forms of model support. NRC staff stated that without this additional model support, it would be difficult to argue the relative likelihood of the base case compared to alternative cases. Additionally, NRC staff indicated that the uncertainty analysis results not be used to demonstrate compliance with the performance objectives because (i) there is limited support for the base case and (ii) there is limited support for the assignment of the likelihood of alternative cases and consequently, the averaging of alternative cases in the “All Cases” model. NRC staff recommended DOE present the results of alternative cases individually and provide qualitative information regarding the likelihood of alternative cases. Finally, NRC staff indicated that DOE should use the results of its probabilistic analysis to inform areas where additional model support is needed.



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6.2 #


Recommendation [FTF and HTF TERs – Next PA Update] In the FTF TER, NRC staff recommended DOE improve transparency and documentation of its benchmarking process. NRC recommends DOE apply a more methodical and systematic approach to the benchmarking process in future updates to its PA. In the HTF TER, NRC staff also noted that DOE could improve its benchmarking process.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions The HTF benchmarking process was detailed in the most recent HTF Goldsim model update (SRR-CWDA-2014-00060), with improved transparency and documentation. This recommendation will be considered further in the next PA revision.

Recommendation [HTF TER – Next PA Update] In the HTF TER, NRC staff noted that due to significant differences between the GoldSim and PORFLOW modeling results, the NRC staff plans to continue to evaluate the PORFLOW modeling assumptions and results for the compliance case (Case A) during the monitoring period to provide confidence that the timing of peak dose is not artificially delayed. This applies to the inadvertent intruder analysis.


Recommendation [FTF TER – Next PA Update] In the FTF TER, NRC staff suggested that DOE consider consistency between the plotting interval and calculation time step size. DOE should also correct errors in its probabilistic assessment (e.g., porosity of 1E-20).


Recommendation [FTF TER – Next PA Update] NRC made a general comment that DOE could improve its parameter distribution assignments, hybrid modeling approach, benchmarking process, and evaluation and interpretation of probabilistic modeling results. With respect to parameter distributions, NRC included several items in its open items database (see Appendix B in ML12212A192), most of which are listed in other recommendations, with the exception of probability of basemat bypass flow.


Recommendation [ML116277A060] As a result of its review of the revised dose methodology for Liquid Waste PAs (SRR-CWDA-2013-00058, Revision 1), NRC identified a number of items that should be addressed in future HTF and FTF PAs revisions. The complete listing of items is documented in the Dose Calculation Methodology TRR.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions These items will be considered in the next PA revision.


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6.3 #

FTF PA Revisions (Additional Considerations)

General NRC Monitoring Activities [ML15238A761] NRC staff will review the revised PAs and issue a TER documenting the results of its review. NRC staff will pay special attention to supporting documentation generated since the last PA revision, including results of experiments, analog studies, models, and peer reviews conducted to support key Monitoring Areas listed in the NRC Monitoring Plan. NRC will evaluate any revisions to the tank farm PAs to ensure inadvertent intrusion into tank farm components were properly evaluated in the 10 CFR 61.42 analyses. This MF can be closed when NRC staff concludes that DOE has adequately evaluated FEPs and scenarios related to inadvertent intrusion in its PA documentation and that its LW PA Maintenance Program is sufficient to evaluate new and significant information related to 10 CFR 61.41 and 10 CFR 61.42 in the future.

Section 3.2.2 – Prepare Out-Year FTF PA Revisions Section 3.2.3 – Prepare Out-Year HTF PA Revisions Future revisions to the FTF PA will be provided to NRC for review in support of NRC’s monitoring role.

No specific NRC recommendations identified. Section – N/A

MA 7 – Protection of Individuals During Operations

7.1 ±

Protection of Workers During Operations

General NRC Monitoring Activities [ML15238A761] NRC staff should review, on at least an annual basis, DOE reports and records that are related to dose during waste disposal operations to assess whether doses are within the limits found in 10 CFR Part 20 and are ALARA. NRC staff should periodically confirm programs and policies presented in the waste determination (DOE/SRS-WD-2012-001) continue to be in effect during the operational period. In particular, NRC staff should verify personnel involved in waste disposal operations are provided dosimetry and are familiar with requirements of the radiation protection program. NRC will leverage staff in its Region I office with experience in radiation protection inspections to support onsite observations in this area. This factor will be closed at the end of the assumed 100-year institutional control period or after operational doses are expected to be reduced to non-risk-significant levels following tank closure activities.


Recommendation [FTF and HTF TERs – Ongoing] No specific recommendations identified, NRC concluded DOE can demonstrate compliance with protection of individuals during operations and DOE provides adequate information that individuals will be protected during operations.

Section – N/A


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7.2 ± Air Monitoring

General NRC Monitoring Activities [ML15238A761] NRC will review air monitoring data to determine whether activity released in the air, as a result of tank farm disposal activities, could cause a MOP located at the SRS boundary to receive a dose of greater than 10 mrem/yr through the air pathway. NRC staff should periodically confirm the air monitoring program continues to adequately assess the risk of tank farm operations. As part of this review, NRC staff should evaluate whether sampling locations and sampling methodologies are adequate to assess airborne emissions from the tank farms or rely on independent verification from the SCDHEC. NRC staff expects the dose from airborne emissions to be small. If the airborne emissions dose becomes more risk significant, then NRC staff will need to evaluate the air monitoring program in greater detail. This factor will be closed at the end of the assumed 100-year institutional control period or when operational doses are expected to be reduced to non-risk-significant levels following tank closure activities.


Recommendation [FTF and HTF TERs – Ongoing] No specific recommendations identified, NRC concluded DOE can demonstrate compliance with protection of individuals during operations and DOE provides adequate information that individuals will be protected during operations.

Section – N/A

7.3 ± ALARA

General NRC Monitoring Activities [ML15238A761] NRC staff’s monitoring of ALARA under 10 CFR 61.43 will be carried out through monitoring of the Radiation Protection Program and related activities. NRC staff should periodically (or at the appropriate time relevant to each measure) review documents associated with the following measures for ensuring ALARA: (i) a documented Radiation Protection Program; (ii) a Documented Safety Analysis; (iii) radiological design for protection of occupational workers and the public; (iv) regulatory and contractual enforcement mechanisms; (v) access controls, training, and dosimetry; and (vi) occupational radiation exposure history. These measures are described in the waste determination or basis documents (DOE/SRS-WD-2012-001 and DOE/SRS-WD-2014-001). This factor will be closed at the end of an assumed 100-year institutional control period or when operational doses are expected to be reduced to non-risk-significant levels following tank closure activities.


No specific NRC recommendations identified. Section – N/A


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MA 8 – Site Stability

8.1 #

Settlement

General NRC Monitoring Activities [ML15238A761] Technical reviews and onsite observations of settlement will be conducted by the NRC staff to assess compliance with 10 CFR 61.44. Reviews will focus on (i) settlement data collected during closure operations of the tank farms, (ii) settlement data collected from analogous sites, and (iii) updated settlement modeling investigations. NRC Technical reviews related to the risk significance of calcareous zones will be conducted to assess compliance with 10 CFR 61.44. Reviews will focus on (i) processes that have resulted in the formation of sinks at the SRS and specifically at the tank farms at the General Separations Area, (ii) the potential for these processes to affect site stability throughout the performance period, and (iii) the potential dose consequences from subsidence related to dissolution of calcareous sediment. DOE stated that it will consider static-loading-induced settlement, seismically induced liquefaction and subsequent settlement, and seismically induced slope instability in the final design of the closure cap. NRC staff will review DOE’s consideration of these processes as information is made available. To assess compliance with 10 CFR 61.44, NRC staff will visually observe the facility for obvious signs of degeneration of the facility. For example, evidence of ponded water on the cap surface may be a sign of differential settlement. Surface fractures may be evidence of underlying displacement. NRC staff also may plan site visits to observe the facility after severe weather events (e.g., storms, tornados) to ascertain how well the facility can withstand these events. NRC staff should also discuss any maintenance activities that are performed at the disposal facility (e.g., repairs to engineered surface barriers) with SCDHEC. This monitoring activity is expected to remain open indefinitely.


Recommendation [ML15238A761] NRC staff expects DOE to inform it of changes to features in the immediate area that might affect site stability. These changes may include (i) vegetation denudation at the surface due to fires or storms; (ii) erosion features caused by extreme precipitation events or long-term processes; or (iii) visible surface changes due to significant biotic intrusion, earthquakes, or other geological processes.


Recommendation [FTF and HTF TERs – Intermediate-Term] In the FTF and HTF TERs, NRC staff recommended DOE continue to evaluate closure cap settlement and stability, including consideration of (i) increased overburden from the tank grout and closure cap on settlement and (ii) potential for subsidence associated with ongoing dissolution of calcareous sediment in the Santee Formation.



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8.1 #

Settlement

Recommendation [FTF and HTF TERs – Intermediate-Term] In the FTF and HTF TERs, NRC staff concluded that assumed long-term compressive strength of the grout monolith is not adequately supported and may be optimistic based on observations of vault cracks, discussed in TER Section 4.2.9.1 (ML112371715). While cracking of the vault concrete and tank grout is not expected to result in significant structural tank collapse, the integrity of the vault concrete and tank grout is important to steel liner performance and waste release.


TER Recommendations Only

N/A # N/A

Recommendation [FTF and HTF TERs] In the FTF TER, NRC staff recommended DOE specifically consider and evaluate HRR removal in its technology selection and effectiveness evaluations consistent with the NDAA. In the HTF TER, NRC staff recommended DOE provide more emphasis on removal of HRR in its technology selection process and provide a clear linkage between the HTF PA results, including information regarding the long-term risks associated with the HTF facility, and the demonstration that HRRs have been removed to the MEP per Criterion 2.


Recommendation [FTF and HTF TERs] In the FTF TER, NRC staff recommended DOE continuously evaluate new technologies, participate in technology exchanges, and not default to previous evaluations for technology selection. In the HTF TER, NRC staff recommended DOE continue evaluating new technologies for future use as tank closure progresses, especially if previously used technologies are no longer practical to use. Furthermore, for those tanks in which conditions are dissimilar (e.g., Tank 48), the NRC staff would expect DOE to reevaluate technologies as opposed to relying on previously performed technology evaluations. The NRC staff also recommended DOE continue its efforts to participate in technology exchanges so that it can stay informed of potential new cleaning technologies. New technologies or improvements to current technologies should be fully considered in the selection process for future tank cleaning. DOE should try to optimize operational parameters for existing technologies and technologies to be developed in the future to ensure that removal of HRRs is not hampered or made more difficult because of poor planning or lack of investment in waste characterization.


Recommendation [HTF TER] In the HTF TER, NRC staff indicated that DOE’s approach to optimization of technology through sampling and monitoring during cleaning should be documented. The NRC staff also recommended DOE consider how it might better assess and optimize the effectiveness of selected technologies (e.g., obtain better baseline information).



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N/A # N/A

Recommendation [HTF TER] In the HTF TER, NRC staff noted that, although the results from mapping contain uncertainties, performing the tank mapping methodology during multiple cleaning phases will provide additional information on the effectiveness of specific technologies. As such, the NRC staff recommended DOE perform the tank mapping consistently and as frequently as practical throughout the cleaning process.


Recommendation [HTF TER] In the HTF TER, NRC staff recommended DOE should obtain better baseline information from which it could better assess oxalic acid effectiveness.


Recommendation [HTF TER] In the HTF TER, NRC staff supported DOE’s efforts to re-evaluate oxalic acid cleaning against downstream impacts to determine the future role of oxalic acid cleaning, as opposed to relying on previous evaluations of oxalic acid technology.


Recommendation [HTF TER] In the HTF TER, NRC staff noted that each final characterization should be accompanied by a Technical Task Request and a Quality Assurance and Quality Control Plan.


Recommendation [HTF TER] In the HTF TER, NRC staff indicated that to help overcome the limitations encountered with cleaning Tanks 5 and 6 for the cleaning of future tanks, the NRC staff recommends that DOE evaluate the effectiveness of the submersible mixer pumps with respect to bulk sludge removal versus residual heel removal. The NRC staff also recommends that DOE compare the efficiency and effectiveness of the submersible mixer pump to previously used technologies or readily available technologies.


Recommendation [HTF TER] In the HTF TER, NRC staff noted that if oxalic acid is not available to be used for cleaning future tanks and a technology with similar proven effectiveness is not used as an alternative, DOE may need to reconsider the validity of assuming that the cooling coil and tank wall surface inventory is negligible.


Recommendation [HTF TER] In the HTF TER, NRC staff recommended DOE should develop separate site-specific factors for risk-significant annular waste versus tank waste sources in the future. Annular and tank sources would then be separately compared to adjusted waste classification concentration limits to determine the classification of HTF components.



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N/A # N/A

Recommendation [HTF TER] In the FTF TER, NRC recommended DOE more fully evaluate or document its consideration of alternatives to additional HRR removal, including (i) modifications to existing technologies (e.g., upgraded Mantis or enhanced chemical cleaning); (ii) modification to tank system components (e.g., installation of new risers or removal of equipment from existing risers); (iii) sequential cleaning (e.g., sequencing of mechanical and chemical technologies in Tank 18F); and (iv) alternative cleaning technologies (e.g., alternative reagents to leach HRRs out of residual heels).


Recommendation [HTF TER] In the FTF TER, NRC recommended DOE better quantify technology effectiveness. For example, DOE should better characterize waste and residual tank inventory prior to deployment of cleaning technologies to better assess effectiveness. In the HTF TER, NRC staff recommended, to the extent practical, DOE consider obtaining data on HRR inventories prior to and following major cleaning campaigns (e.g., before and after treatment of Type I tanks with oxalic acid) to provide effectiveness measurements for chemical cleaning and mechanical feed-and-bleed.


Recommendation [HTF TER] In the HTF TER, NRC staff noted that given the potential risk significance of the waste remaining in the Tank 16 annulus, the NRC staff recommends that DOE more fully evaluate the practicality of additional radionuclide removal from the Tank 16H annulus versus the long-term benefit of reduced risk considering uncertainty in the releases of radionuclides from the Tank 16 annulus. While DOE’s HTF PA demonstrates that the risk from waste remaining in the annulus is reasonable, alternative waste release models may lead to higher risk estimates. NRC staff went on to note that at this stage, DOE has provided a rough order of magnitude cost-benefit analysis of additional HRR removal from the Tank 16H annulus to the NRC staff (U-ESR-H-00107). The NRC staff acknowledges that DOE is still preparing the final removal report and recommends that DOE provide a more detailed cost benefit analysis to support the Criterion 2 demonstration for Tank 16H in the final removal report. NRC staff indicated that it would like to obtain a copy of the final removal report when it is complete.

Section 3.1.3 – Provide General Technical Support on Tank Farm PA Issues Ongoing support of NRC’s monitoring role will be provided as necessary. An SA specific to the HTF Type I and Type II Tanks was prepared and issued in August 2016. [SRR-CWDA-2016-00078] This SA updated the radiological and chemical inventories for the HTF Type I and Type II tanks, incorporating lessons learned from the final waste tank characterization results to date.

Recommendation [HTF TER] In the HTF TER, NRC staff noted that DOE improved the operating plan for Tank 12H by requiring the availability of the transfer receipt tank to be confirmed prior to acid addition. The NRC staff encourages DOE to continue to analyze the lessons learned from these prior cleaning campaigns to prevent limitations of the liquid waste system from unexpectedly influencing the effectiveness of future cleaning campaigns.



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N/A # N/A

Recommendation [FTF and HTF TERs] In the FTF TER, the NRC staff recommended DOE include more specificity in its process for determining HRRs are removed to the MEP, including (i) defining the term end states versus removal goals and (ii) clarifying when conditions are sufficiently similar to warrant use of a previous technology evaluation. NRC staff also recommended DOE continue to better define the documented process to be used to demonstrate removal to the MEP to ensure consistent (nonarbitrary) application of the criterion. In the HTF TER, NRC staff noted that Appendix B of the draft basis for the waste determination for HTF (DOE/SRS-WD-2013-001) outlines a general approach to demonstrate that the HRRs will be removed to the MEP. However, DOE could still improve the standardization of metrics for determining that the anticipated end states have been reached.


1 Monitoring Factors are color-coded based on NRC-determined prioritizations in NRC Monitoring Plan. [ML15328A761] Symbols are included for clarity. § High Priority Recommended * Lower Priority ‡ High Priority Dependent or More Difficult ± Not Prioritized - Periodic † Medium Priority # Not Prioritized – Not Periodic

2 Recommendations noted by “[FTF and HTF TER – timing]” are from the NRC TERs for FTF [ML112371715] and HTF [ML14094A496]. Other recommendations are from the FTF and HTF Monitoring Plan Table A-1 [ML15328A761], which replaced the FTF Monitoring Plan [ML12212A192], or from other various NRC TRRs. [ML12272A082, ML12272A124, ML13273A299, ML13085A291, ML13269A365, ML13277A063, ML13100A230, ML13080A401, ML14342A784, ML15301A710, ML15301A830, ML16196A179, ML16231A444, ML16277A060]